ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the Idaho National 

Engineering and Environmental Laboratory (INEEL)

Site Description

Document Number:

ORAUT-TKBS-0007-2 

Effective Date: 11/07/2003

Revision No.: 00

Controlled Copy No.: _______

Page 1 of 70

Subject Expert: Norman Rohrig

Document Owner

Approval: __Signature on File__________Date:11/06/2003

Norman Rohrig, TBD Team Leader

Approval: ___ Signature on File ________Date: 11/06/2003

Judson L. Kenoyer, Task 3 Manager

Concurrence: __ Signature on File _____Date: 11/06/2003

Richard E. Toohey, Project Director

Approval: _____ Signature on File ______Date: 11/07/2003

James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

LIST OF TABLES ................................................................................................................................

Page

4

LIST OF FIGURES..............................................................................................................................4

Record of Issue/Revisions ...................................................................................................................5

Acronyms and Abbreviations ...............................................................................................................6

2.1 Introduction...............................................................................................................................12

2.2 Test Area North.........................................................................................................................14

2.2.1 Technical Support Facilities........................................................................................... 14

2.2.1.1 Hot Shop/Warm Shop/Hot Cells Annex, 1955 to present........................................14

2.2.1.2 TAN 607 Storage Pool 1955 to present..................................................................16

2.2.1.3 Storage Pads TAN 690 and TAN 691.....................................................................16

2.2.1.4 Radioactive Liquid Waste Disposal System 1959 to present ..................................17

2.2.1.5 Radioactive Parts Service and Storage Area 1955 to present ................................17

2.2.1.6 TAN 607 Radiography Facility................................................................................17

2.2.2 Initial Engine Test Facility, 1955 to 1966 ....................................................................... 18

2.2.2.1 Heat Transfer Reactor Experiment No.1, December 27, 1955 to January 3, 

1959.......................................................................................................................18

2.2.2.2 Heat Transfer Reactor Experiment No. 2, July 1957 to March 28, 1961 .................18

2.2.2.3 Heat Transfer Experiment No. 3, 1958 to December 1960 .....................................19

2.2.2.4 Systems for Nuclear Auxiliary Power 10A Transient No.1, Early 1960s..................19

2.2.2.5 Systems for Nuclear Auxiliary Power 10A Transient No. 3, April 1, 1964 ...............20

2.2.2.6 Systems for Nuclear Auxiliary Power 10A Transient No. 2, 1965 to January 

11, 1966.................................................................................................................20
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2.2.3 Water Reactor Research Test Facility, 1958 to 1973..................................................... 20

2.2.3.1 Shield Test Pool Facility .........................................................................................21

2.2.3.2 Low Power Test Facility (LPTF), 1958 to 1973.......................................................21

2.2.3.2.1 Critical Experiment Tank, 1958 to 1960 ..........................................................21

2.2.3.2.2 Hot Critical Experiment, 1958 to March 28, 1961 ............................................22

2.2.3.2.3 Split Table Reactor System, 1971...................................................................22

2.2.3.2.4 Fast Spectrum Refractory Metals Reactor (710), March 1962 to 1968 ............22

2.2.3.2.5 Cavity Reactor Critical Experiment, May 17, 1967, to Early 1970s..................22

2.2.3.2.6 Spherical Cavity Reactor Critical Facility, 1972 to November 1973 .................23

2.2.3.2.7 High Temperature Marine Propulsion Reactor (630-A), 1962 to 1964.............23

2.2.4 Loss of Fluid Test Facility, 1973 to July 9, 1985 ............................................................ 23

2.2.5 Specific Manufacturing Capability (SMC), 1985 to present ............................................ 24

2.3 Idaho Nuclear Technology and Engineering Center, November 1951 to present ...................... 24

2.3.1 Fuel Processing Facility, February 1953 to 1992........................................................... 25

2.3.2 INTEC Fuel Storage Facility (CPP-603), 1950 to present .............................................. 27

2.3.3 Fluorine Dissolution Process and Fuel Storage Facility, 1984 to present....................... 28

2.3.4 High Level Liquid Waste Underground Storage Tanks (Tank Farm), 1951 to 

present..........................................................................................................................28

2.3.5 Waste Calcining Facility-1 and New Waste Calcining Facility, December 1963 

To 2002.........................................................................................................................29

2.3.6 INEEL Comprehensive Environmental Response Compensation and Liability 

Act Disposal Facility Complex, July 2003 to present...................................................... 30

2.3.7 TMI-2 Independent Spent Fuel Storage Installation – CPP-1774 to present.................. 30

2.4 Argonne National Laboratories-West, February 1951 to present............................................... 30

2.4.1 Experimental Breeder Reactor No. I, April 1951 to December 30, 1963 ........................ 31

2.4.2 Boiling Water Reactor Experiment No. 1, Late 1953 to July 22, 1954............................ 31

2.4.3 Boiling Water Reactor Experiment No. 2, October 19, 1954, to March 1955.................. 32

2.4.4 Boiling Water Reactor Experiment No. 3, June 9, 1955, to December 1956 .................. 32

2.4.5 Boiling Water Reactor Experiment No. 4, December 3, 1956, to June 1958 .................. 32

2.4.6 Boiling Water Reactor Experiment No. 5, February 9, 1962 to September 1964............ 33

2.4.7 Zero Power Reactor No. 3, October 1955 to November 1970........................................ 33

2.4.8 Argonne Fast Source Reactor, October 29, 1959, to Late 1970 .................................... 34

2.4.9 Transient Reactor Test Facility, February 23, 1958, to April 1994.................................. 34

2.4.10 Experimental Breeder Reactor No. II, May 1961 to September 30, 1994....................... 34

2.4.11 Hot Fuel Examination Facility, 1964 to 1968.................................................................. 35

2.4.12 Zero Power Physics (Plutonium) Reactor, April 18, 1969 to April 1992 (standby).......... 36

2.4.13 Neutron Radiography Facility, October 1, 1977 to present ............................................ 36

2.4.14 Fuel Assembly and Storage Building, 1970 to present................................................... 37

2.4.15 Other ANL-West Support Facilities ................................................................................ 37

2.5 Radioactive Waste Management Complex, May 1952 to present ............................................. 38

2.5.1 Subsurface Disposal Area ............................................................................................. 38

2.5.2 Intermediate Level Transuranic Storage Facility............................................................ 38

2.5.3 Transuranic Storage Area ............................................................................................. 39

2.5.4 Stored Waste Examination Pilot Plant ........................................................................... 39

2.5.5 Transuranic Package Transporter Loading Station........................................................ 39

2.5.6 Advanced Mixed Waste Treatment Project.................................................................... 39

2.6 Central Facilities Area, August 2, 1943, to present.................................................................... 39
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2.6.1 The Hot Laundry............................................................................................................ 40

2.6.2 The Health Physics Instrument Laboratory.................................................................... 40

2.6.3 DOE Laboratory Accreditation Procedure Irradiation Facility ......................................... 40

2.6.4 Radiological Environmental Science Laboratory............................................................ 41

2.6.5 CF-674 Building............................................................................................................. 41

2.7 Test Reactor Area, MARCH 31, 1952, to present...................................................................... 41

2.7.1 Materials Test Reactor, March 31, 1952, to April 23, 1970 ............................................ 41

2.7.2 Engineering Test Reactor, September 19, 1957, to December 1981............................. 42

2.7.3 Advanced Test Reactor, July 7, 1967, to present .......................................................... 42

2.7.4 Reactivity Measurement Facility, February 11, 1954, to April 10, 1962.......................... 43

2.7.5 Advanced Reactivity Measurement Facility No.1, October 10, 1960, to 1974 ................ 43

2.7.6 Advanced Reactivity Measurement Facility No. 2, December 14, 1962 to 1968 ............ 43

2.7.7 Coupled Fast Reactivity Measurement Facility, 1968 to 1991........................................ 44

2.7.8 Engineering Test Reactor Critical Facility, May 20, 1957, to 1982 ................................. 44

2.7.9 Advanced Test Reactor Critical Facility, May 19, 1964 to present ................................. 44

2.7.10 TRA Hot Cell Facility 1954 to present............................................................................ 45

2.7.11 TRA Gamma Facility, 1955 to unknown......................................................................... 45

2.7.12 Radiation Measurements Laboratory, 1952 to present .................................................. 45

2.7.13 Radiochemistry Laboratory............................................................................................ 46

2.7.14 Liquid Waste Disposal Ponds ........................................................................................ 46

2.7.15 High-Level Liquid Waste Tanks and Transfer Facility .................................................... 46

2.8 Auxiliary Reactor Area, April 1958 to late 1990s ....................................................................... 47

2.8.1 ARA-I ............................................................................................................................47

2.8.2 ARA-II/SL-1, February 23, 1960, to April 6, 1961........................................................... 47

2.8.3 ARA-III........................................................................................................................... 48

2.8.4 ARA-IV, March 30, 1961 to May 29, 1964 ..................................................................... 48

2.9 Waste Reduction Operations Complex/Power Burst Facility/Special Power Excursion 

Reactor Test Areas ...................................................................................................................48

2.9.1 SPERT-I, June 11, 1955 to 1964................................................................................... 49

2.9.2 SPERT-II, March 11, 1960, to October 1964 ................................................................. 49

2.9.3 SPERT-III, December 19, 1958, to June 1968............................................................... 49

2.9.4 SPERT-IV, July 24, 1962, to August 1970..................................................................... 50

2.9.5 Power Burst Facility, September 22, 1972, to 1985 ....................................................... 50

2.9.6 Lead Storage Facility..................................................................................................... 50

2.9.7 Waste Experimental Reduction Facility, 1982 to 2001................................................... 50

2.9.8 Mixed Waste Storage Facility ........................................................................................ 51

2.10 Organic Moderated Reactor Experiment, September 17, 1957, to April 1963 ........................... 51

2.11 Test Grid III, 1957 to April 24, 1970........................................................................................... 52

2.11.1 Fuel Element Burn Tests ............................................................................................... 52

2.11.2 Fission Products Field Release Tests, July 25, 1958, to September 26, 1958............... 52

2.11.3 Relative Diffusion Tests, November 30, 1967, to October 1, 1969................................. 53

2.11.4 Experimental Cloud Exposure Study, May 3, 1968, to April 24, 1970 ............................ 53

2.12 Experimental Dairy Farm/Experimental Field Station ................................................................ 53

2.12.1 Controlled Environmental Radioiodine (Release) Tests, May 27, 1963, to 

December 1977............................................................................................................. 53
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2.13 INEEL Research Center, 1984 to present ................................................................................. 54

2.14 Army Reentry Vehicle Facility Site (or Station), 1965 to May 1996............................................ 54

References ........................................................................................................................................58

Glossary ............................................................................................................................................62

LIST OF TABLES

Table Page

2.1. INEEL prime site contractors.................................................................................... 13/56

2.2 Other operating facility technical area contractors .................................................... 13/56

2.3 Radionuclides of concern for all reactors....................................................................... 56

2.4 ICPP 601/602 process cell information.......................................................................... 57

LIST OF FIGURES

Figure

1 Map of the Idaho National Engineering and Environmental Laboratory ......................... 15

Page
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RECORD OF ISSUE/REVISIONS

ISSUE

AUTHORIIZATION

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 08/31/2003 00-A Original issue draft technical basis document for the 

Idaho National Engineering and Environmental 

Laboratory (INEEL) Site Description. Initiated by 

Norman Rohrig.

Draft 10/02/2003 00-B Draft revision to incorporate NIOSH and internal 

review comments. Initiated by Norman Rohrig.

11/07/2003 11/07/2003 00 First approved issue. Initiated by Norman Rohrig.
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-TKBS-0007-2 Page 6 of 70

ACRONYMS AND ABBREVIATIONS

α alpha particle

AEC Atomic Energy Commission

AECL Administrative Exposure Control Level

AEDE annual effective dose equivalent

AFSR Argonne Fast Source Reactor

AL Analytical Laboratory

AMWTP Advanced Mixed Waste Treatment Program

ANC Aerojet Nuclear Corporation

ANL Argonne National Laboratory

ANL-E Argonne National Laboratory East

ANL-W Argonne National Laboratory West

ANP Aircraft Nuclear Propulsion

ANPP Aircraft Nuclear Propulsion Program

APS Atmospheric Protection System

ARA Army Reactor Area (later Auxiliary Reactor Area)

AREA Army Reactor Experimental Area

ARMF-1 Advanced Reactivity Measurements Facility No.1

ARMF-2 Advanced Reactivity Measurements Facility No.2

ARVFS Army Reentry Vehicle Facility Site

ATR Advanced Test Reactor

ATRC Advanced Test Reactor Critical Facility

β beta particle

BNFL British Nuclear Fuels Limited

BORAX Boiling Water Reactor Experiment 

BORAX I Boiling Water Reactor Experiment No. 1

BORAX II Boiling Water Reactor Experiment No. 2

BORAX III Boiling Water Reactor Experiment No. 3

BORAX IV Boiling Water Reactor Experiment No. 4

BORAX V Boiling Water Reactor Experiment No. 5

BWR Boiling Water Reactor

C Celsius/centigrade

CA Contamination Area

CAM continuous air monitors

CAMU Corrective Action Management Unit

CAS Criticality Accident System

CDC Capsule Driver Core

CE Critical Experiment

CEL Chemical Engineering Laboratory

CEDE committed effective dose equivalent

CERCLA Comprehensive Environmental Response Compensation and Liability Act

CERT Controlled Environmental Radioiodine Tests (later Radionuclides)

CET Critical Experiment Tank

CFA Central Facilities Area

CFRMF Coupled Fast Reactivity Measurement Facility

c/m counts per minute

CPP Chemical Processing Plant

CRCE Cavity Reactor Critical Experiment
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CTF Core Test Facility

D&D decontamination and decommissioning

DAC Derived Air Concentration

DE dose equivalent

DL detection limit

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DU depleted uranium

DOE-ID DOE-Idaho Operations Office

DOELAP DOE Laboratory Accreditation Program

EBR Experimental Breeder Reactor

EBR-I Experimental Breeder Reactor No.1

EBR-II Experimental Breeder Reactor No.2

EBOR Experimental Beryllium Oxide Reactor

EBWR Experimental Boiling Water Reactor

ECCS Emergency Core Cooling System

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EFS Experimental Field Station

EOCR Experimental Organic Cooled Reactor

EPA Environmental Protection Agency

ERDA Energy Research and Development Administration

ETF Effluent Treatment Facility

ETR Engineering Test Reactor

ETRC Engineering Test Reactor Critical 

EXCES Experimental Cloud Exposure Study

F Fahrenheit

FARET Fast Reactor Test

FASB Fuel Assembly and Storage Building

FAST Fluorinel Dissolution Process and Fuel Storage Facility

FCF Fuel Cutting Facility

FDF Fluorinel Dissolution Facility

FEBT Fuel Element Burn Tests

FFTF Fast-Flux Test Facility

FMF Fuel Manufacturing Facility

FPF Fuel Processing Facility

FPR Fuel Processing Restoration

FPFRT Fission Product Field Release Test

FSF Fuel Storage Facility

γ gamma

GCRE Gas-Cooled Reactor Experiment

GE General Electric Corporation

HCA High Contamination Area

HEPA high-efficiency particulate air

HEU highly enriched uranium

HFEF Hot Fuel Examination Facility

HLLW high-level liquid waste
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HLW high-level waste

HOTCE Hot Critical Experiment

HP health physicist/health physics

HPIL Health Physics Instrument Laboratory

HPP Hot Pilot Plant

HRA High Radiation Area

HSL Health Services Laboratory

HTRE Heat Transfer Reactor Experiment

IBM International Business Machines

IC Initial Criticality

ICDF INEEL CERCLA Disposal Facility

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

ID Idaho

IDO Idaho Operations Office

IWP Industrial Waste Pond

IFSF Irradiated Fuel Storage Facility

IET Initial Engine Test

IFR Integral Fast Reactor

ILTSF Intermediate-Level Transuranic Storage Facility

IMBA Internal Modular Bioassay Analysis

INC Idaho Nuclear Corporation

INL Idaho National Laboratory

INEC Idaho Nuclear Energy Commission

INEL Idaho National Engineering Laboratory

INEEL Idaho National Engineering and Environmental Laboratory

INTEC Idaho Nuclear Technology and Engineering Complex

IRC INEEL Research Center

IREP Interactive RadioEpidemiological Program

ISU Idaho State University

ISFSI Independent Spent Fuel Storage Installation

IWP Industrial Waste Pond

keV Kiloelecton volt

kVp Kilovolt peak

kW kilowatt

LANL Los Alamos National Laboratory

LLD lower limit of detection

LMFBR Liquid Metal Fast Breeder Reactor

LOCA Loss-of-Coolant Accident

LOFT Loss-of-Fluid Test

LPTF Low-Power Test Facility

LSC Liquid Scintillation Counter

MDF Materials Development Facility

MeV megaelectron volt

ML-1 Mobile Low-Power Reactor

MPBB Maximum Permissible Body Burden
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MPLB Maximum Permissible Lung Burden

MPOB Maximum Permissible Organ Burden

mR milliroentgen

mrem millirem

ms millisecond

MTR Materials Testing Reactor

MW megawatt

NaK Sodium potassium

NASA National Aeronautics and Space Administration

NCRP National Council on Radiation Protection and Measurements

NERP National Environmental Research Park

Nf neutron flux-fast

NIOSH National Institute for Occupational Safety and Health

NOAA National Oceanic and Atmospheric Administration

NOCTS NIOSH OCAS Claims Tracking System

NPR New Production Reactor

NRAD Neutron Radiography Facility

NRC Nuclear Regulatory Commission

NRF Naval Reactors Facility

NRTS National Reactor Testing Station

NTA Nuclear Track Emulsion

Nth neutron flux-thermal

NWCF New Waste Calcining Facility

OCAS (NIOSH) Office of Compensation Analysis and Support

OMRE Organic Moderated Reactor Experiment

ORAU Oak Ridge Associated Universities

PA Posterior-Anterior (medical X-ray)

PAS Personal Air Sampler

PBF Power Burst Facility

PCB polychlorinated biphenyl

PIC Pocket Ionization Chamber (i.e., "pencil" dosimeter)

PIF Process Improvement Facilities

psi pounds per square inch

PWR pressurized-water reactor

R Roentgen

rad radiation absorbed dose

RAF Remote Analytical Facility

RAL Remote Analytical Laboratory

RaLa radioactive lanthanum

RAM Radiation/Remote Area Monitor

RBE Relative Biological Effectiveness

RBOF Receiving Basin for Offsite Fuel

RCRA Resource Recovery and Conservation Act

RCT Radiological Control Technician

RDT Relative Diffusion Test

rem Roentgen equivalent man

RESL Radiological Environmental Sciences Laboratory
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rep Roentgen equivalent physical

RHA Radiation Hazards Analysis

RLWTF Radioactive Liquid Waste Treatment Facility

RMF Reactivity Measurement Facility

RML Radiation Measurements Laboratory

RPSSA Radioactive Parts Service and Storage Area

RU recycled uranium

RWMC Radioactive Waste Management Complex

RWP Radiological Work Permit

RSAC Radiological Safety Analysis Computer

RSWF Radioactive Scrap and Waste Facility

SCRCE Spherical Cavity Reactor Critical Experiment

SDA Subsurface Disposal Area

SEC Special Exposure Cohort

SID Source to Image Distance

SIS Special Isotope Separations

SL-1 Stationary Low-Power Reactor

SM-1 Stationary Medium-Power Reactor

SMC Specific Manufacturing Capability

SNAP Systems for Nuclear Auxiliary Power

SNAPTRAN Systems for Nuclear Auxiliary Power Transient

SPERT Special-Power Excursion Reactor Test

SPF Sodium Processing Facility

SRS Savannah River Site

SSD Source to Skin Distance

SSSTF Staging, Storage, Sizing, and Treatment Facility

STM Stack Tritium Monitor

STEP Safety Test Engineering Program

STP Sewage Treatment Ponds

STPF Shield Test Pool Facility

STR split-table reactor

SWDF Solid Waste Disposal System

SWEPP Stored Waste Examination Pilot Plant

SWP Safe Work Permit

t thermal

Τ1⁄2 Half life of a nuclide or isotope

TAN Test Area North

TBD Technical Basis Document

TEPC Tissue Equivalent Proportional Counter

TLD Thermoluminescent dosimeter

TLND Thermoluminescent neutron dosimeter

TMI Three Mile Island

TRA Test Reactor Area

TREAT Transient Reactor Test

TRIGA Training Research and Isotope General Atomic

TSA Transuranic Storage Area

TRU transuranics

TRUPACT Transuranic Packaging Transporter

TSF Technical Support Facility
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USGS U. S. Geological Survey

WAC Waste Acceptance Criteria

WBRR Western Beam Research Reactor

WCA Waste Characterization Area

WCF Waste Calcining Facility

WERF Waste Experimental Reduction Facility

WIPP Waste Isolation Pilot Plant

WROC Waste Reduction Operations Complex

WRRTF Water Reactor Research Test Facility

ZPR Zero Power Reactor

ZPR III Zero Power Reactor No.3

ZPPR Zero Power Plutonium Reactor (later Zero Power Physics Reactor)
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2.1 INTRODUCTION

The purpose of this document is to provide an Idaho National Engineering and Environmental 

Laboratory (INEEL) “Site” profile that contains technical basis information for use by the Oak Ridge 

Associated Universities (ORAU) Team, or any dose reconstruction action, to evaluate the total 

individual occupational dose for claimants under the Energy Employees Occupational Illness 

Compensation Program Act (EEOICPA). 

The INEEL played a major role in early reactor research and development. The Site has operated 52 

reactors plus fuel handling and reprocessing and radioactive waste storage and disposal facilities 

since it began operations in 1949. The purpose of this Technical Basis Document (TBD) is to assist in 

the evaluation of worker dose from INEEL processes using the methodology in NIOSH OCAS-IG-001 

External Dose Reconstruction Implementation Guideline (NIOSH 2002a) and OCAS-IG-002 Internal 

Dose Reconstruction implementation Guideline (NIOSH 2002b). 

This TBD provides supporting technical data to evaluate, with claimant-favorable assumptions, the 

total INEEL occupational dose that can be reasonably associated with worker radiation exposure as 

covered under EEOICPA legislation. The documentation in this TBD addresses evaluation of 

monitored and unmonitored worker exposure and missed dose. In addition, it presents the technical 

basis of methods used to prepare the INEEL worker dose records for input to the NIOSH Interactive 

RadioEpidemiological Program (IREP) and the Internal Modular Bioassay Analysis (IMBA) computer 

codes used to evaluate worker dose. It also presents information on the uncertainty for recorded 

INEEL exposure and dose values.

This section describes the facilities and processes and historical information related to worker internal 

and external exposures for use when actual monitoring data may be unavailable. 

Proving the Principle, A History of the Idaho National Engineering and Environmental Laboratory, 

1949-1999, by Susan M. Stacy (Stacy 2000) was used as a resource in the development of facility 

and process descriptions.

INEEL or “the Site” is an 890-square-mile reservation encompassing almost 572,000 acres with a 

maximum distance of about 39 miles from north to south and 36 miles from east to west. It is 30 to 60 

miles west of Idaho Falls, Idaho. Major Site-related research facilities and offices are in Idaho Falls. 

The Site, situated on the Snake River Plain of southeastern Idaho at an elevation of about 5,000 ft., is 

above the Snake River Plain Aquifer. 

In 1949, the U.S. Atomic Energy Commission (AEC) established the National Reactor Testing Station 

(NRTS) in Idaho as a Federal reservation to build, test, and operate nuclear reactors. The Site

utilized a variety of support facilities and equipment. In 1974, the NRTS became the Idaho National 

Engineering Laboratory (INEL) and, in 1997, the Idaho National Engineering and Environmental 

Laboratory (INEEL). In 2004, the Site will become the Idaho National Laboratory (INL). 

INEEL is unique among U.S. Department of Energy (DOE) facilities because it is a large complex site 

with many independent technical areas, contractors, goals, and missions. Table 2-1 lists historical 

prime Site contractors and the years they operated. The prime contractor provided some services to 

the other contractors and operated most of the facilities. Some of the specific technical areas were 

operated in part by other contractors during the tenure of the prime contractor. Table 2-2 lists those 

facilities and technical areas and the dates contractors other than the prime contractor operated them.
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Table 2-1. INEEL prime site contractors. 

Operating years Prime contractor

1950–1966 Phillips Petroleum Company

1966–1972 Idaho Nuclear Corporation (Allied Chemical Corporation, Aerojet General Corporation, 

and Phillips Petroleum Company)

1972–1976 Aerojet Nuclear Corporation

1976–1994 EG&G Idaho

1994–1999 Lockheed Martin Idaho Technologies Company

1999–2004 Bechtel BWXT Idaho, LLC

Table 2-2. Other operating facility technical area contractors. 

Argonne National Laboratory–West

1949–2004 University of Chicago

Idaho Chemical Processing Plant (Idaho Nuclear Technology and Engineering Center)

1950–1953 American Cyanamid Company

1953–1966 Phillips Petroleum Company

1966–1971 Idaho Nuclear Corporation

1971–1979 Allied Chemical Corporation

1979–1984 Exxon Nuclear Idaho Company

1984–1994 Westinghouse Idaho Nuclear Company

Aircraft Nuclear Propulsion Program (ANPP)

1952–1961 General Electric

Specific Manufacturing Capability

1983–1986 Exxon Nuclear Idaho Company

1986–1991 Rockwell INEL

1991–1994 Babcock & Wilcox Idaho, Inc.

Dosimetry services at INEEL were unique among DOE facilities in that DOE Idaho Operations Office 

(DOE-ID) personnel operated and provided internal and external dosimetry services. The DOE-ID 

dosimetry branch provided and analyzed external dosimetry badges, counted workers in the whole 

body counter, and analyzed bioassay samples. Originally, DOE-ID personnel at the Radiological 

Environmental Sciences Laboratory (RESL) in Building CF-690 in Central Facilities Area maintained 

exposure histories of personnel based on dosimetry records, including bioassay data. With the 

advent of the DOE Laboratory Accreditation Program (DOELAP), dosimetry responsibility was 

transferred to the prime contractor on January 2, 1989, to eliminate a conflict of interest on the part of 

DOE-ID. 

Figure 2-1 shows the relative location of each facility or technical area discussed in this TBD. The 

following sections describe each facility and its processes, with the exception of the Naval Reactors 

Facility (NRF), a naval propulsion facility exempted under EEOICPA. The subsections for the facilities 

and technical areas and processes contain information on the particular area. 

Personnel working at INEEL in designated radiological areas were typically required to wear state-ofthe-art dosimetry (film badges, thermoluminescent dosimeters [TLDs], personal ion chambers [PICs], 

respiratory protection, anticontamination clothing, etc.). Facilities and radiological areas were 

monitored by remote and portable radiation/remote area monitors (RAMs) and continuous air monitors 

(CAMs). Portable RAMs, CAMs, air samplers, etc., were used for work where fixed units were not 

available. In cases where airborne radioactivity might be present or where internal exposure was 

possible, applicable respirators were provided to prevent/reduce internal exposure. 
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Engineered systems were incorporated as practicable to minimize the potential for radiological 

airborne radioactivity and direct radiation exposure. Bioassay programs were instituted to monitor 

and assess potential internal exposures. Environmental monitoring systems were placed around the 

Site at multiple locations to measure direct radiation, fallout, effluent discharges or releases to onsite 

personnel and to members of the public. Each facility has had film or TLD badges in specific building 

areas and around perimeter fences to measure direct environmental radiation accumulation at the 

location as a check and balance on source terms within Site locations. 

In addition to the nuclear reactor experiments discussed below in relation to the TRA, there have been 

other such experiments at INEEL. Table 2-3 lists common radionuclides for reactors.

2.2 TEST AREA NORTH

TAN is 30 miles northeast of the Central Facilities Area. General Electric Corporation (GE) opened 

TAN in 1952 for the Aircraft Nuclear Propulsion (ANP) Program, which operated during the 1950s and 

early 1960s. Additional facilities built to support the program were the Initial Engine Test (IET) 

Facility, the Technical Support Facility (TSF), and the Water Reactor Research Test Facility (WRRTF)

During the 1970s the Loss of Fluid Test (LOFT) Facility was built at the west end of the TAN area next 

to the aircraft hanger built to support the ANP Program. LOFT was a smaller version of a commercial 

pressurized-water reactor designed to allow, create, or recreate loss-of-fluid accidents. In 1983, 

construction started on the classified Project X, which later became the Specific Manufacturing 

Capability (SMC). The SMC, which was inside the aircraft hanger, manufactured depleted uranium 

(DU) armor for the M1-A1 Abrams Main Battle Tank.

2.2.1

TSF was built to provide technical and administrative support for the ANP Program. The facilities 

included a large Hot Shop, hot cells, storage pool, cafeteria, machine shop, office space, etc. A fourrail railroad system, with a shielded locomotive and two turntables, connected the IET facility to the 

Hot Shop, the Warm Shop, and eventually the LOFT facility. 

Technical Support Facilities

2.2.1.1 Hot Shop/Warm Shop/Hot Cells Annex, 1955 to present

The TAN Hot Shop is in the TAN-607 building. It is 51 feet wide by 165 feet long by 55 feet high. The 

walls are 7 feet thick and the windows are 6 feet thick to provide protection to personnel involved in 

the examination, handling, analysis, or disassembly of radioactive components.

The Warm Shop, immediately adjacent to the Hot Shop, is one large open room measuring 51 feet 

wide by 80 feet long by 50 feet high. A four-rail track system connects the Warm Shop to the Hot 

Shop and supports TAN projects. Experiments, projects, or equipment with relatively low-level 

contamination or direct radiation were brought to the Warm Shop for modification or repair.

The Hot Cell annex is on the south side of the Hot Shop and consists of a hot cell and control 

galleries. The interior of the Hot Cell is 10 feet wide by 35 feet .long. The Hot Cell Annex (TAN-633) 

is a one-story building north of and adjoining the storage pool. It consists of four shielded cells with 

interlocking sliding drawers for transferring samples. 

 
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Figure 2-1. Map of the Idaho National Engineering and Environmental Laboratory. 

Facilities include the Argonne National Laboratory-West (ANL-W), where the 

Experimental Breeder Reactor No. 2 (EBR-II), Transient Reactor Test Facility (TREAT) 

and Zero Power Plutonium Reactor (ZPPR) are located; Auxiliary Reactor Area (ARA); 

Central Facilities Area (CFA); Idaho Chemical Processing Plant (ICPP) [now the Idaho 

Nuclear Technology and Engineering Complex (INTEC)]; Experimental Breeder Reactor 

No. 1 (EBR-I); Boiling Water Reactor Experiment (BORAX); Radioactive Waste 

Management Complex (RWMC); Special Power Excursion Reactor Test (SPERT) area; 

Grid III, the test grid where the Fuel Element Burn Tests (FEBT) A and B occurred; Test 

Reactor Area (TRA); the Experimental Field Station (EFS); Naval Reactors Facility; Test 

Area North (TAN), where the Initial Engine Tests (IETs) occurred; and the Core Test 

Facility (CTF) at TAN.
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The Hot Shop and Hot Cells are equipped with cranes, manipulators, and other equipment for remote 

handling and work on experiments of all types. The Hot Shop was used extensively for refueling and 

repairs on the Heat Transfer Reactor Experiment (HTRE) reactors. Many other activities involving 

radiation levels measured up to 50 rad hr-1 for brief periods have occurred in the facilities, including 

disassembling the Stationary Low-Power Reactor (SL-1) and receiving, examining, and storing Three 

Mile Island (TMI) fuel and debris. 

Internal exposure potential exists from airborne radioactivity of mixed fission products and activation 

products associated with reactor operation (Table 2.1.3). 

External exposure

2.2.1.2 TAN 607 Storage Pool 1955 to present

 exists from mixed fission products and mixed activation products (MAP). Radiation 

levels varied from background levels to measured levels greater than 50 rad hr-1, principally gamma 

rays with energy greater than 250 keV. 

The TAN Storage Pool is adjacent to the north side of the Hot Shop. It is 70 feet long, 48 feet wide, 

and 24 feet deep. An underground passageway (19 feet by 24 feet) under the Hot Shop north wall 

connects the main pool to the Hot Shop vestibule. The vestibule, in the northeast corner of the Hot 

Shop, is 25 feet long, 24 feet wide and 24 feet deep. The top of the passageway under the shield wall 

is 5 feet under water to protect the main pool area from radiation sources in the Hot Shop.

The TAN Pool was used for storage of reactor fuel, including some commercial fuel and TMI core 

debris. In mid-2002, the TMI fuel and debris were moved to INTEC for storage. Radioactive 

materials being put in the pool or removed are generally transferred in shielded casks. The casks can 

be transferred via the vestibule in the Hot Shop or loaded directly on a truck in the north end of the 

pool. The pool is not lined and does not meet current criteria for radioactive storage pools (Bonney et 

al. 1995). 

The water in the Storage Pool is contaminated with mixed fission (MFP) products, activation products 

and transuranics. The major isotopes in the pool are 137Cs, 147Pm, 241Pu, 241Am, and 90Sr, with an 

average water concentration of 4.6 X 10 -4 μCi ml-1 (Bonney. et al. 1995).

Internal exposure potential exists from airborne radioactivity primarily from the nuclides listed above. 

External exposure

2.2.1.3 Storage Pads TAN 690 and TAN 691

exists from the nuclides listed above. 

The Storage Pads are near the main east-west railroad west of the Hot Shop and east of the 

turntable. The TAN 690 pad was used to store three “Abnormal Waste Casks.” Two of the casks are 

empty; the third contains filter elements with activity greater than or equal to 100 nCi g-1 of 

transuranics. External radiation levels are less than or equal to 0.1 mrem hr-1. 

Storage Pad TAN-691 is west of TAN 607 and contains the “Spent Fuel Storage Casks.” These 

casks were designed to investigate the feasibility of storing spent nuclear fuel from commercial 

reactors in a dry state without external cooling requirements. The casks have fuel from a Virginia 

Power reactor and from LOFT. The casks, which are filled with helium gas and sealed, are monitored 

for temperature and pressure and alarm if either deviates from accepted limits. Measured radiation 

levels near the casks are 25 to 30 mrem hr-1 gamma and about 40 mrem hr-1 neutron. The casks are 

periodically surveyed and assessed for radioactive contamination. 
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Internal exposure potential exists from potential airborne leaks. 

External exposure

2.2.1.4 Radioactive Liquid Waste Disposal System 1959 to present

exists from activities proximate to the casks. Not all personnel wear neutron 

dosimetry. 

A radioactive liquid waste system (TAN 666), which was built in 1959, collected and processed 

intermediate-level liquid wastes generated in the TAN area and transferred them to one of three 

underground 10,000-gallon, stainless-steel collection tanks (V1, V2, and V3). The liquid wastes were 

concentrated in an evaporator in TAN 616, and the concentrated solution was pumped to one of two 

50,000-gallon underground liquid waste feed tanks (V9 and V10). Solids were separated and sent to

the RWMC. Originally, the liquid effluent was combined with low-level radioactive liquid waste and 

discharged to a disposal well. In 1972, INEEL replaced the disposal well with a disposal pond. The 

pond was an unlined diked area encompassing about 35 acres that could receive about 33 million 

gallons per year (ERDA 1977). From 1959 to 1974, TAN reported liquid effluent releases to the 

disposal well or pond of 58 curies, with highest releases in 1959, 1968, and 1969. 

Internal exposure exists from potential airborne mixed fission products and activation products. 

External exposure

2.2.1.5 Radioactive Parts Service and Storage Area 1955 to present

exists from activities around the liquid waste system. 

Storage buildings (TAN-647/TAN-648) and the adjacent storage pads are known as the Radioactive 

Parts Service and Storage Area (RPSSA). The facility is in the northwest portion of the TAN TSF 

area. The area has residual contamination from earlier projects, including the Heat Transfer Reactor 

HTRE-2 and HTRE-3 reactors. The buildings and the contaminated areas are marked with perimeter 

fencing and warning signs. Some of the radiologically contaminated soil in the area resulted from 

movement of SL-1 debris into the Hot Shop, and some particles were measured as high as 50 rad hr-1

beta near contact (90Sr/Y). The gates and buildings are locked when access is not required.

The RPSSA buildings and areas on and around the pads contain casks, boxes, and equipment of all 

types that is either contaminated or radioactive and has to be in a controlled storage area.

Internal exposure potential exists from airborne radioactivity of leaking packages or disturbing 

contaminated soil areas containing mixed fission products and/or activation products. 

External exposure

2.2.1.6 TAN 607 Radiography Facility

 comes from several items in storage that have had radiation levels in the R hr-1

range from mixed fission products and activation products. 

The TAN 607 Radiography Facility was in TAN 607, south of the TAN Warm Shop. The facility used 

60Co and 192Ir sources and X-ray units for nondestructive examination of metal welds, parts, or 

equipment. Some of the isotopic sources were in the 100-curie range when first purchased. 

Radiography was performed by trained and certified technicians according to approved procedures. 

Periodic radiography would be required on a reactor or piece of equipment that could not be brought 

to the facility. Applicable procedures were required to evaluate radiological hazards and establish 

necessary controls. The Radiography Facility is not operational at present.
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Internal exposure is negligible for radiographic activities in the Radiography Facility.

External exposure

2.2.2

occurred from activities associated with radiographic sources. 

The IET (TAN-620) was approximately 1 mile north of the TAN TSF area. It was built as a shielded 

test facility to prove that heat from a nuclear energy source could run a turbojet engine. The IET had 

high-density reinforced concrete walls 2 feet thick and ceilings 3 feet thick. The floor of the facility 

was 15 feet below grade with 14 feet of dirt over the top. After a HTRE test, a locomotive driver in a 

shielded cab would hook onto the reactor and return it to the TAN TSF Hot Shop. There were a total 

of 26 IET runs involving three separate reactor assemblies – HTRE 1, 2, and 3. Jet engines were 

fitted to a HTRE reactor at TSF and transported by the four-rail system to the IET facility.

Initial Engine Test Facility, 1955 to 1966

During the early 1960s, the AEC initiated another nuclear safety program at IET, called Systems for 

Nuclear Auxiliary Power (SNAP) Transient (SNAPTRAN). The program evaluated the hazards 

associated with using nuclear reactors for aerospace auxiliary power systems. In addition, it was 

designed to investigate the consequences of a nuclear accident. Three SNAPTRAN tests were 

conducted, with the last two ending in destruction of the SNAP 10A/2 reactors.

The various test series involved a number of power reactor operations that resulted in the release of 

radionuclides to the environment. The dates and times of reactor operations, and consequent 

releases, varied throughout the duration of the project. Therefore, the meteorological conditions that 

existed depended on the time of the tests. All operations were under strict meteorological control, and 

permissible wind directions seriously limited operations. On many days, it was impossible to operate 

at all, and most of the time operation was possible only a few hours each day (Thornton, Rothstein, 

and Culver 1962).

Internal exposure was possible from airborne mixed fission products and activation products. It was 

minimized by a filtered intake air system that kept the air pressure in the occupied areas positive in 

relation to outside air. The exhaust from the HTRE jet engines was vented up a 150-foot stack. 

External exposure

2.2.2.1 Heat Transfer Reactor Experiment No.1, December 27, 1955 to January 3, 1959

occurred from the activation and mixed fission product inventory after tests were 

completed and personnel worked on the associated test assemblies. 

On November 4, 1955, the HTRE-1 reactor was placed on the test pad at the IET and brought to 

power. On December 30, 1955, the reactor was again brought to power with a J47 turbojet engine 

attached. The reactor was run on the test stand for 150.8 hours at full power, 20 MW thermal, 

exceeding the design requirement of 100 hours. During the first 6 hours of full-power operation, fuel 

element damage occurred in three cartridges caused by a defect in the insulation liners. After the 

damaged elements were replaced, power operation resumed. The test was successful in proving the 

reactor could run the engines without chemical fuel. The water-moderated and water-cooled reactor 

used enriched uranium fuel clad in nickel-chromium (Thornton, Rothstein, and Culver 1962).

2.2.2.2 Heat Transfer Reactor Experiment No. 2, July 1957 to March 28, 1961

The HTRE-2 "parent core" was similar to the HTRE-1 core except the central seven air tubes were 

removed and replaced by a hexagonal void 11 inches across the flat. A corresponding opening was 

made in the top shield plug so sections of advanced reactors could be inserted in the parent core 
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without requiring removal of the core from the shield. This converted it to a materials test reactor, 

which subjected test fuels to environments reaching 2,800°F for extended periods and higher 

temperatures for short periods. The ANP Program advanced the technology of high-heat ceramic 

reactor fuels. HTRE-2 operated for 1,299 hours at powers up to 14 MW. Test sections consisted of 

metallic fuel elements combined with air-cooled hydrided zirconium moderators and beryllium oxide 

fuel elements for use in ceramic reactors (Thornton, Rothstein, and Culver 1962; Stacy 2000).

2.2.2.3 Heat Transfer Experiment No. 3, 1958 to December 1960

A new HTRE engine was developed with the reactor, engine shielding, and heat transfer system 

arranged in a horizontal configuration anticipating final design in an airframe. HTRE-3 operated for 

126 hours, running two modified J47 jet engines at power as high as 32 MW thermal. This reactor 

was water-cooled and water-moderated, and used uranium fuel clad in nickel-chromium. In 

December 1960, HTRE-3 ran two turbojet engines at 2000°F without the help of any chemical fuel 

(Thornton, Rothstein, and Culver 1962).

On November 18, 1958, a nuclear excursion occurred during the IET No. 13 test. The excursion was 

the result of the dynamic and shim rods being withdrawn by the control system under the influence of 

an erroneous reactor power indication. Activity was released from the exhaust stack and a narrow 

band of fallout was contained fully within the boundaries of the NRTS. The maximum dose rate 

observed in the Assembly and Maintenance area and approximately 3,000 feet from the cloud 

centerline was .04 mrem hr-1. The maximum fallout observed, at about 4 hours after the incident, 

measured 0.8 to 2.0 mrem hr-1 at contact roughly 1 1/2 miles from IET (DOE 1991a).

After each HTRE test, the cooling water was drained and replaced with mercury to provide shielding 

to personnel working on the reactors. After the HTRE program ended, HTRE 2 and 3 reactors were 

parked in the RPSSA. The HTRE engines were moved to the EBR-1 historic site in 1988.

Internal exposure potential existed from the airborne release inventory during reactor operation as 

argon and other constituents in the cooling air became radioactive. Fuel elements occasionally 

ruptured, which released mixed fission products and activation products to the outdoor environment. 

In some cases the cooling air flow to fuel elements was deliberately blocked to determine fuel failure 

parameters and characteristics. Fourteen of the tests were categorized as group 1 releases under 

National Council on Radiation Protection and Measurements (NCRP) methodology (NCRP 1996); 

tests 10 and 4 were ranked the highest. Seven tests were ranked group 2. Fission Product 

inventories were based on documented reactor operating histories (RAC 2002). Operations and 

support personnel were inside the pressurized control room (TAN-620) during reactor operation 

during the testing phase of the HTRE reactor program minimizing their exposure potential. .

External Exposure

2.2.2.4 Systems for Nuclear Auxiliary Power 10A Transient No.1, Early 1960s

occurred from the mixed fission product and activation product inventory during 

work associated with test assemblies after tests were completed. 

The AEC initiated a safety program to evaluate hazards associated with using nuclear reactors for 

aerospace power systems. The portion of the program concerned with determining the kinetic 

behavior of the SNAP 10A/2 reactors and the consequences of certain nuclear accidents involving 

these reactors was called SNAPTRAN. The tests were conducted at the IET facility at TAN. The 

SNAPTRAN program extended the SPERT reactor safety testing program (see Section 2.9) to 

aerospace applications. Three test series involving three reactors investigated the behavior of 

SNAP10A/2 fuel under large-transient, power-excursion conditions. SNAPTRAN-1 was subjected to 
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nondestructive tests in conditions approaching but not resulting in damage to the zirconium-hydrideuranium fuel. The SNAP 10A/2 reactors were 9 inches in diameter by 12 inches long and were 

composed of a sodium/potassium (NaK)-cooled core containing 37 rods of fully enriched uranium in a 

zirconium-hydride matrix.

2.2.2.5 Systems for Nuclear Auxiliary Power 10A Transient No. 3, April 1, 1964

SNAPTRAN-3 was the first of two destructive tests on a version of the small reactor (SNAP10A/2) 

designed to supply auxiliary power in space. The test, conducted at the IET Facility on April 1, 1964, 

simulated the accidental fall of a reactor into water or wet earth, as could occur during assembly, 

transport, or launch abort. The test demonstrated that the reactor would destroy itself immediately 

instead of building up a high inventory of radioactive fission products. The test involved reactor 

operation at a power level of 30,000 MW for 1.5 milliseconds (ms). More than 99% of the fission 

products inventory was retained in the surrounding water and reactor fuel remains. No airborne 

iodine was detected, so it was presumed that halogens were retained in the water as well as 

particulate radionuclides (Cordes et al. 1967). 

2.2.2.6 Systems for Nuclear Auxiliary Power 10A Transient No. 2, 1965 to January 11, 1966

This test version of the small space reactor, SNAP 10A/2, was intentionally destroyed on January 11, 

1966. It provided information on the dynamic response, fuel behavior, and inherent shutdown 

mechanisms of these reactors in an open-air environment. In normal operation, the control drums of 

the SNAP10A/2 would be rotated to obtain criticality after the reactor was in orbit. In a launch abort, 

however, impact on the earth might cause the reactor drums to rotate inward and the core to go 

critical and conceivably destroy itself, releasing fission products to the surrounding environment. The 

test data contributed to an understanding of reactor disassembly on impact and methods for 

assessing or predicting the radiological consequences. The reactor core was 93% enriched fuel, 

containing 4.75 kg of 235U. The reactor operated at a power level of 36,000 MW for 1.5 ms. The 

fission product release fractions were reported to be 0.75 for noble gases, 0.70 for iodines, 0.45 for 

tellurium, and 0.04 for solids. Reactor operation was assumed to generate 4681 curies of 41Ar. 

SNAPTRAN meteorological conditions for testing were strict. Weather requirements were to consist 

of lapse conditions with no rainfall, and were to persist a minimum of 3 hours after the tests. Wind 

had to be southwest (180 ̊ to 240 ̊) between 10 and 30 mph (Cordes et al.1965).

Internal exposure potential during the SNAP 10A test series was well controlled by requiring 

personnel to stay inside the IET during tests. During the SNAPTRAN 2 and 3 tests, the reactors were 

totally destroyed, releasing fission products including 131I, all of which was tracked. 

External exposure

2.2.3

to mixed fission products and mixed activation products was received during 

cleanup of the reactor debris, which was scattered around the test pad and along the four-track rail 

system. Dosimetry was required for all personnel. 

The WRRTF is approximately 1.25 miles southeast of the TSF area. The first facility constructed at 

WRRTF was the Low Power Test Facility (LPTF) in 1958. This facility was also known as Semiscale 

and was in the east quadrant of WRRTF. The Shield Test Pool Facility (STPF), in the west quadrant 

of WRRTF, was built as part of the ANP Shielding Experimentation Program; in 1963, the pool facility 

was modified for the Experimental Beryllium Oxide Reactor (EBOR). The EBOR project was 

cancelled in 1966 before construction was complete. 

Water Reactor Research Test Facility, 1958 to 1973
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The Semiscale facility in LPTF was a forerunner of the LOFT. It was a scaled mockup of one loop of 

a four-loop pressurized-water reactor (PWR). The facility was electrically heated to provide steam to 

run blowdown tests (Shaw, Boucher, and Loomis undated). No radiological exposure resulted from 

the blowdown tests.

Due to the electrically heated experiment cores in the facilities listed in the subsections of 2.2.3, 

internal and external exposure from neutron reactions in the reactors did not exist. Exposures will be 

noted as applicable at the end of the pertinent test series.

2.2.3.1 Shield Test Pool Facility

The STPF, initially used for a reactor experiment known as Susie, was used for bulk shielding 

experiments performed in support of the ANP Shielding Experimentation Program. Susie utilized a 

sample canister box that was pressurized with inert gas or air to keep it dry. It was at the reactor 

centerline and contained 16 sample tubes in which organic samples were irradiated. Susie was a 

swimming-pool-type reactor; water-moderated, water-cooled, water-reflected, and shielded by 

approximately 17 feet of water. The fuel loading was 4 kg of 235U and the nominal power level was 2 

MW thermal. After the ANP Program ended in 1961, Susie was used by other programs at the NRTS 

(Walsh 1961). The facility would later become known as EBOR. Since the EBOR project was 

cancelled, other experiments were conducted using tracer-level nuclides to investigate the reaction of 

water in piping, as well as instrument calibrations. The cell area was filled with piping for pressurized 

water tests using up to 25-curie 137Cs sources in a radiography-type environment in which the source 

material was cranked out of a shielding cask to be in the proximity of the piping. The Cs and tracer 

sources have been removed. 

Internal exposure at the STPF was not known to occur due to the low power of Susie, the water 

environment, and use of sealed small sources. 

External exposures

2.2.3.2 Low Power Test Facility (LPTF), 1958 to 1973

occurred from the use of sealed sources, primarily 137Cs. 

The LPTF was used to conduct several low-power (less than 100-watt) reactor research programs. 

The LPTF contained two shielded cells with three independent control rooms and necessary support 

facilities. The north cell, Room 101, was called the Critical Experiment (CE) cell and the south cell, 

Room 102, was the Initial Criticality (IC) cell. The test cells are of poured concrete construction with a 

4-foot-thick wall between them. The walls between the cells and the control room are 5 feet thick for a 

height of 30 feet. The outside wall of the IC cell is 2 feet thick, and the outside wall of the CE cell is 3 

feet thick for a height of 30 feet (Kunze and Chase 1970). 

The construction of the facility was such that more than one reactor program could be running at the 

same time. Heavy experiment pieces could be moved in or out of each cell through large rollup doors 

in the back. 

2.2.3.2.1 Critical Experiment Tank, 1958 to 1960

The Critical Experiment Tank (CET) reactor was part of the ANP Program in the CE Cell of the LPTF. 

The CET was a low-power reactor originally designed to mock-up the HTRE-1 and HTRE-2 reactors. 

It was used primarily to perform critical experiments for insert tests in the HTRE-2 power plant. Fuel 

test bundles intended for testing in HTRE-2 were first evaluated for reactivity characteristics in the 

CET. The water-moderated CET utilized a beryllium reflector (Becar et al. 1961).
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CET was one of three low-power reactors supporting the ANP Program, along with the STPF Reactor 

(Susie) and the Hot Critical Experiment (HOTCE) (Hoefer 1957). The ANP Program ended in 1961.

2.2.3.2.2 Hot Critical Experiment, 1958 to March 28, 1961

Located in the LPTF CE cell, HOTCE was an elevated-temperature critical experiment designed to 

obtain information on temperature coefficients of solid moderated reactors. The fuel elements 

consisted of fuel bearing stainless-steel wire 1/8 inch diameter. The maximum loading was 50.4 kg of 

93.2% enriched UO2. The reactor used a hydrided zirconium moderator and a beryllium reflector. 

The hexagonal prism-shaped core and reflector were mounted such that the fuel cells were horizontal. 

One half was mounted on a fixed table and the other in a movable table so the two halves were 

separated. The normal operating power was 1 watt for a period of 1 to 3 hours. The reactor could be 

operated at 100 watts for short periods (Hoefer 1957). 

2.2.3.2.3 Split Table Reactor System, 1971

The purpose of the Split Table Reactor System was to provide nuclear information on a variety of fast 

and thermal spectrum reactors. The reactors were to be assembled, operated, and revised to perform 

experiments with both thermal and fast systems. The reactor was operated in the CE cell of the 

LPTF.

The reactor was a split-table type, 7 ft 2 in wide and 11 ft long. The reactor was opened and closed 

using a hydraulic system. The table was an aluminum matrix structure composed of a stack of 

hexagonal tubes mounted horizontally on each table half. When the two halves were brought 

together, a single reactor assembly was formed. Normal operating power level was 0 to 500 watts 

thermal not to exceed 1000 watts or 10 kilowatt hours per month (Lofthouse 1971).

2.2.3.2.4 Fast Spectrum Refractory Metals Reactor (710), March 1962 to 1968

The 710 reactor was a split-table, low-power critical facility at LPTF. The objective was to collect data 

for a proposed fast-spectrum, refractory-metal reactor concept called the 710 Reactor. The concept 

involved using metals such as tungsten and tantalum in a compact, very-high-temperature reactor for 

generating power in space.

2.2.3.2.5 Cavity Reactor Critical Experiment, May 17, 1967, to Early 1970s

The Cavity Reactor Critical Experiment (CRCE) was an outgrowth of a program started by the 

National Aeronautics and Space Administration (NASA) at LPTF to investigate the propulsion of 

space rockets by nuclear power, offering the possibility of much greater thrust per pound of propellant 

than chemical rockets. The concept for the cavity reactor core was that the uranium would be in a 

vapor, or gaseous state. Hydrogen propellant flowing around it would theoretically attain much higher 

temperatures (up to 10,000°F) than in conventional solid-core rockets. The experiments at TAN used 

simulated hydrogen propellant and produced data on the reactor physics feasibility of a gaseous core 

being able to go critical. The core was uranium hexafluoride (UF6); the experiments were performed 

at the relatively low temperature of about 200°F. In the proposed ultimate application, the ball of 

uranium gas would have been held in place by the hydrogen flowing around it, something like a pingpong ball suspended in a stream of air. Uranium core temperatures as high as 100,000°F were 

considered possible. During the CRCE program, the UF6 was always contained in the storage vessel 

or the reactor cavity.
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2.2.3.2.6 Spherical Cavity Reactor Critical Facility, 1972 to November 1973

The Spherical Cavity Reactor Critical Facility (SCRCE) was the final experiment in reactor physics 

work for the NASA-sponsored program to determine the feasibility of a reactor going critical with a 

gaseous core of UF6. Previous work used a cylindrical configuration because of its ease of 

construction. The spherical shape was considered a more likely geometry for the ultimate application 

in a rocket to Mars. The SCRCE assembly consisted of two aluminum tanks, one inside the other, 

with D2O in the space between the two tanks. The D2O would act as a reflector and moderator and, 

during normal shutdown, would be transferred from the reactor to a storage tank (INC 1969).

2.2.3.2.7 High Temperature Marine Propulsion Reactor (630-A), 1962 to 1964

The 630-A reactor was a low-power critical experiment operated at the LPTF. The mission of the 630-

A was to explore the feasibility of an air-cooled, water-moderated system for nuclear-powered 

merchant ships. Development ended in December 1964 after decisions to lower the priority of the 

entire Nuclear Power Merchant Ship Program.

Internal exposure was not known to occur. 

External exposures

2.2.4

occurred during cell entry after a reactor run. 

The LOFT reactor at TAN 650 was a centerpiece in the safety testing program for commercial power 

reactors. The reactor was a scale model of a commercial pressurized-water power plant built to 

explore the effects of loss-of-coolant accidents (LOCAs). Thirty-eight nuclear power tests were 

conducted with various accident scenarios, including the accident at TMI. Among other goals, the 

program investigated the capability of emergency core cooling systems to prevent core damage 

during a LOCA. Experiments at LOFT simulated small-, medium-, and large-break LOCAs, 

sometimes complicated with other events such as “loss of offsite power.” LOFT was deactivated in 

1986, following completion of the LP-FP-2 experiment, the most significant severe fuel damage test 

ever conducted in a nuclear reactor. That test, which involved the heating and melting of a 100-rod 

experimental fuel bundle, provided information on the release and transport of fission products that 

could happen during an actual commercial reactor accident where core damage occurred.

Loss of Fluid Test Facility, 1973 to July 9, 1985

The LOFT facility was built in a steel domed reactor containment vessel, 97 feet high, with a 

basement. The LOFT Control and Support Building is four stories high with a basement. The 

structures are attached at the basement level, and both have reinforced concrete exterior walls. 

The control room, visitor center, experimental data recording and display area, sample counting area, 

and operation support room were in TAN 630, a two-story underground building. During the final 

preparations for a test, the containment vessel doors were closed and the only access to the facility 

was through a shielded underground tunnel. During reactor operation and testing, personnel were 

restricted to safe facilities in TAN 630.

Internal exposure was possible from airborne fission product activity in the containment soon after 

shutdown. Entries were monitored with a CAM, and respiratory protection was worn as required. On 

July 9, 1985, following completion of the LOFT LP-FP-2 test, leakage was discovered from the fission 

product monitoring system and the primary coolant system, allowing fission products to enter the 

reactor building. Over the following 2-month period, 8,780 curies of noble gas (88 Kr) and 0.09 curie of 

Iodine (131I) were released to the environment (Hoff, Chew, and Rope 1986; Stachew 1985).
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External exposure

2.2.5

occurred to personnel working inside the containment vessel or on the primary 

system or sample systems associated with the reactor. During initial entry after a test, the fields in 

containment were ≥ 100 mrad hr-1 beta gamma. The short-lived fission products would die off rapidly, 

reducing the general fields to ≤ 10 mrad hr-1 beta gamma. 

The SMC project is in the ANP Program aircraft hanger (TAN 629) and surrounding buildings. The 

project consists of Phase I, Phase II, and support facilities. A Materials Development Facility (MDF), 

located in TAN 607A and the TAN Hazardous Waste Storage Area, located in TAN-628, were 

formerly part of the SMC project, but have been decommissioned and turned over for other use. The 

SMC project is classified. In 1991, the mission was declassified in that the SMC manufactures armor 

for the Army made out of depleted uranium (DU). The major radioactivity in the DU is 238U, 234Th, 

234mPa, and 234U.

Specific Manufacturing Capability (SMC), 1985 to present

The SMC facilities consist of Phase I in TAN 629 and Phase II in TAN-679 and TAN-681. Metal 

fabrication activities are performed in Phase I facilities. Phase II facilities perform metal rolling in 

TAN-679 and waste processing in TAN-681. SMC formerly used a nitric acid system and this waste 

was processed in TAN-681. This system was removed and replaced with an aqueous system. The 

aqueous waste is processed in TAN-681. All radioactive aqueous waste is collected in storage tanks 

for treatment through an evaporator system and the remaining aqueous waste is solidified and 

disposed of as low-level radioactive waste.

Internal exposure occurred from normal operations using the DU processes. Metal fabrication is the 

primary source of airborne radiological activity, followed by a paint coating process. 

External exposure

2.3 IDAHO NUCLEAR TECHNOLOGY AND ENGINEERING CENTER, November 1951 to

present

occurred through working with the billets of DU. The large pieces were primarily 

handled remotely to minimize exposure. 

INTEC, formerly the Idaho Chemical Processing Plant (ICPP), commonly known as the “Chem Plant”, 

is 53 miles west of Idaho Falls and occupies 200 acres in the middle of the INEEL reservation. The 

Plant stored and processed spent nuclear fuel from university and test reactors all over the world,

from commercial power plants, from most of the reactors at INEEL, and from U.S. Department of 

Defense (DOD) projects. INTEC received the first fuel shipment in November 1951. The first hot run 

started in February 1953. Reprocessing continued until the fuel reprocessing project was cancelled in 

1992. 

The primary INTEC mission involved reprocessing spent nuclear fuel from DOD projects, which 

entailed extracting reusable uranium from spent fuels. Each cladding (e.g., aluminum, zirconium, 

stainless steel) or fuel type called for a different process. The numerous fuel recycling processes 

required support facilities for fuel dissolution and recovery of fissionable materials (e.g., 22 different 

process cells) and waste processing. An innovative high-level liquid waste (HLLW) treatment process 

known as calcining was developed at INTEC as an additional major mission. Calcination reduced the 

volume of liquid radioactive waste generated during reprocessing and placed it in a more stable 

granular solid form.
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In addition to calcination and fuel reprocessing operations, INTEC is a major fuel storage facility 

(FSF). INTEC houses a HLLW evaporation facility, HLLW storage (Tank Farm), airborne radioactive 

waste processing, non-radioactive liquid waste disposal, and the Remote Analytical Laboratory (RAL). 

Although fuel reprocessing ended in 1992, INTEC continues to support other nuclear projects. For 

example, chemical research continues to improve fuel recovery processes, spent nuclear fuel is safely 

stored and prepared for shipment to an offsite repository, development of technology to safely treat 

high-level and liquid radioactive waste that resulted from reprocessing spent fuel continues, past 

environmental releases are being remediated, and some facilities have been decontaminated and 

decommissioned. Inactive INTEC facilities are being evaluated for D&D.

Table 2-3 lists the primary long-lived nuclides used in the INTEC processes. 

Internal exposure might have occurred when workers were near breached and or leaking systems 

containing mixed fission products, transuranics, activation products, etc. 

External exposure 

2.3.1

occurred during maintenance work, laboratory work, fuel cutting, and other support 

work which briefly exposed workers to radiation rates from background levels to measured levels of ≥ 

50 rad hr-1 beta/gamma. 

The INTEC Fuel Processing Facility (FPF) 601/602 processing corridors were used for the chemical 

separation of highly enriched uranium (HEU) from dissolved spent fuel during reprocessing and to 

solidify the recovered HEU for shipment off the site. The process dissolved the fuel in acid, producing 

uranyl nitrate and nitrates of fission products and some transuranics. Solvent extraction separated 

uranium from the fission products. The final product was uranyl nitrate free of impurities and fission 

products. The uranyl nitrate was transferred to the product and shipping area in “Z” Cell through a 

dedicated piping transfer system. Table 2-4 lists the cells and processes associated with the fuel 

reprocessing. With the exception of the Health Physics field office in V cell, cells were accessible only 

during shutdown periods for maintenance and decontamination activities.

Fuel Processing Facility, February 1953 to 1992

In 1956, a process for the recovery of radioactive 140La (RaLa) was assigned to ICPP. The RaLa 

process took place in the complicated “L” Cell, and lasted into 1963. As freshly irradiated Materials 

Testing Reactor (MTR) fuel was dissolved in acid, the dissolving process liberated gases, one of 

which was 131I. 

The high-level liquid waste was transferred and stored in large underground tanks east of the 

processing facility in the area known as the “tank farm.” 

New fuels presented special problems in reprocessing. The Process Improvement Facilities (PIF) and 

the Hot Pilot Plant (HPP) provided information to improve these processes.

The PIF consists of a laboratory building with office space for the technical group. New ideas in 

reprocessing are developed and investigated at the laboratory bench scale. The laboratories were 

designed to handle as much as 1 curie of radioactive materials from Table 2-3 per laboratory. The 

waste stream discharged to the HPP waste system.

The HPP was used to test unproven equipment and systems. The facility consisted of three cells and 

associated utilities. The partition between two of these cells could be removed, making one large cell. 
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The shielding around the cells was equivalent to that of the main plant and provided sufficient 

radiation shielding to run plant-level radioactive material.

The HPP waste system consists of three-level storage. High-level waste in storage can be routed to 

permanent storage tanks. Intermediate-level waste can be routed either to the waste evaporator 

system or the low-level waste tanks. Low-level waste is monitored and discharged to the disposal 

well downstream from the main service waste monitoring system. A service waste monitoring system 

is provided for this secondary stream.

Part of the processing included a Rare Gas Processing Facility (CPP-604). Its purpose was to 

recover 85Kr from spent fuels. This gas product was shipped to Oak Ridge for commercial sale for 

use primarily in leak detection. INTEC was the only source of 85Kr outside the former Soviet Union 

(Cohen et al. 1994). In the late 1950s, the Rare Gas Processing Facility recovered 140Ba for its 140La 

daughter products and shipped it to Los Alamos National Laboratory (LANL) for use in weapon 

research projects.

A decontamination facility in CPP-601 provided support for cleaning tools and equipment for INTEC 

and other INEEL facilities. Items such as water pit gates, cooling pumps, vehicles, etc., were 

decontaminated in this facility. It was also used as a morgue and autopsy facility following the SL-1 

accident. Radiation levels to 25 rem hr-1 open window beta gamma were experienced for brief 

periods in the decontamination facility.

The RAL in CPP-601 has been used for remote examination of hazardous and radioactive materials 

to support INTEC operations. It continues to support INEEL and INTEC activities. The RAL handles, 

processes, analyzes, and experiments on hazardous and radioactive materials of all types. Samples 

collected in stations in the CPP-601 sample corridor were transferred to the RAL for analysis. 

Frequent samples during fuel reprocessing, taken in 10-ml sample bottles, had radiation readings of 

greater than or equal to 25 rad hr-1. During the analytical process the samples were diluted to reduce 

personnel exposure. 

When reprocessing was discontinued in 1992, these facilities were flushed to remove uranium and 

hazardous materials and placed in a standby condition. The Fuel Processing Restoration (FPR) 

Project would have replaced these facilities. FPR was about 40% complete when construction 

stopped in 1992. FPR was discontinued in a manner that preserves the facility for possible use in 

future missions at INTEC.

Three criticality events occurred during FPF operation:

• Criticality Accident of October 16, 1959

•

. A bank of storage cylinders containing a uranium 

solution was air-sparged (air was bubbled violently into the solution to mix it). The cylinders 

were geometrically safe, but the sparging initiated a siphon that transferred 200 liters of the 

solution to a 5,000-gallon tank containing about 600 liters of water. The resulting criticality 

lasted about 20 minutes. No workers were exposed to gamma or neutron radiation because 

the criticality occurred in an unoccupied below-ground cell. Airborne activity spread through 

the plant through vent lines and drain connections, triggering alarms and an evacuation. Two 

people who evacuated received beta exposures as they passed the evacuation route area 

where radioactive gas was being released into the room from floor drains. 

Criticality Accident of January 25, 1961. About 40 liters of uranyl nitrate solution (200 grams of 

uranium per liter) was forced upward from a 5-inch-diameter section of an evaporator into a 

24-inch-diameter vapor disengagement cylinder, well above normal solution level. Analysts
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later assumed that air entered associated lines while operators were attempting to clear a 

plugged line and improve a pump. When the air bubble reached the evaporator, solution was 

expelled from the lower section, and a momentary criticality occurred in the upper section. 

Radiation triggered alarms, but no personnel received ≥100 mrem exposure (Stacy 2000). 

Concrete shielding walls surrounded the location of the criticality; the vent system prevented 

airborne activity from entering work areas; and equipment design prevented a persistent 

excursion. 

• ICPP Criticality Accident of October 17, 1978. A criticality event occurred in the first-cycle 

tributylphosphate extraction system in the CPP-601 process building at the ICPP. The 

incident resulted in no personnel injury, no onsite or offsite contamination, and no damage to 

equipment or property, and the plume traveled over uninhabited areas to the southwest of the 

site (Casto 1980). The criticality caused approximately 3 × 1018 fissions of 235U. The 

Atmospheric Protection System (APS) at INTEC, which became operational in 1975, 

significantly reduced particulate emissions, and filtered all releases associated with the 

criticality event. 

Internal exposure potential existed at the FPF and its support facilities from work) with radioactive 

airborne particulates during maintenance activities, piping/valve changes, the criticality accidents, and 

stack releases. 

External exposure

2.3.2

occurred from work performed in Radiation Areas (RAs) and High Radiation Areas 

(HRAs) containing the nuclides referenced in Table 2-3. Brief exposures to measured levels of 50 rad 

hr-1 existed in the cells during piping/valve changes, decontamination, maintenance activities, working 

production samples in the laboratories, etc.

The original fuel storage facility at INTEC CPP-603 included a special fuel-storage building with three 

20-ft-deep storage pools for spent nuclear fuel. The facility is about a third of a mile south of the main 

processing building. 

INTEC Fuel Storage Facility (CPP-603), 1950 to present 

In addition, CPP-603 contains the Irradiated Fuel Storage Facility (IFSF), a graphite fuel storage area, 

and a Fuel Cutting Facility (FCF). The IFSF stores dry fuel that is incompatible with underwater 

storage. The IFSF has 636 storage positions and is more than half full. Most of the spent fuel stored 

there came from the Fort St. Vrain commercial reactor in Colorado. Shipments from Fort St. Vrain 

ended in 1991.

In the 1950s, the ICPP received a request to process spent fuel elements from Savannah River Site 

(SRS) reactors. The 14-foot elements were clad in aluminum and had to be cut to 18-inch lengths to 

fit in the dissolver vessel. Irradiation of the fuel changed metallurgical characteristics, so instead of 

cutting like regular aluminum they crumbled, necessitating development of a new technique and 

procedure. Equipment change and maintenance was extremely difficult because of the crumbling, 

which contributed to increased levels of exposure and contamination. As a result, complete 

modification of the process and equipment was required.

The pools in CPP-603 were built in 1950 and served as the primary spent fuel storage facility until 

1984. Fuel, once in underwater storage at Building 603, has been transferred to the newer 

underwater storage pools at the Fluorinel Dissolution Process and Fuel Storage Facility (FAST) or to 

dry storage. The facility is being evaluated for decontamination and decommissioning (D&D). 
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Internal exposure potential was greatest from fuel cutting and shearing which created airborne 

radioactivity from the cut SRS fuel elements. 

External exposure

2.3.3

was received during fuel loading and unloading, movements to various locations in 

the storage pools, fuel cutting, cleanup activities, modifications, and D&D evaluations.

FAST has two parts; a spent fuel storage area and the Fluorinel Dissolution Facility (FDF). The 

storage area consists of six storage pools for storing spent nuclear fuel. The FDF includes a hot cell 

with 6-foot-thick concrete walls where spent fuel was dissolved in an acid solution. When 

reprocessing ended in 1992, uranium and hazardous materials were flushed from the FDF, and this 

part of the facility was placed in a standby condition.

Fluorine Dissolution Process and Fuel Storage Facility, 1984 to present

Internal exposure potential existed from airborne radioactivity that may have occurred from the 

various processes associated with the FAST.

External exposure

2.3.4

occurred from unloading and loading irradiated fuel and fuel element examination 

under water and from work in the FDF hot cell environment which created radiation exposure up to 50 

rad hr-1 for brief periods

The HLLW Tank Farm includes 11 underground stainless-steel, 300,000-gallon storage tanks nested 

in a concrete vault east of process building CPP-601. The tanks were used to store radioactive liquid 

waste generated during the reprocessing of spent fuel and plant decontamination work. All of the tank 

farm liquid has been calcined, reducing the volume and converting it to a more stable solid form. The 

underground tanks are encased in concrete vaults that have sumps and leak detection. One tank was 

always kept empty for use as a transfer backup should a problem develop with one of the other tanks. 

The tanks are extremely corrosion-resistant. No leakage has been detected from the tanks. 

However, some leaks have occurred from transfer lines outside the tanks.

High Level Liquid Waste Underground Storage Tanks (Tank Farm), 1951 to present

High-level waste at INEEL is composed of acidic liquid and calcined solids. The acidic liquids have 

been stored in the underground tanks and included actual high-level waste as well as sodium-bearing 

waste that is managed as high-level waste. The stainless-steel tanks allowed the storage of waste in 

acidic form and resisted corrosion. When full, each tank contained only a few gallons of pure 

radioactive fission products. The rest of the solution was dissolved cladding-metal ions, process 

additives, and water. The tanks that received waste from the first cycle extraction, which accumulated 

most of the fission products, had cooling systems to carry away decay heat to minimize corrosion. 

The HLLW Processing Facility, CPP-604, is east of CPP-601. Liquid waste generated from 

reprocessing activities was transferred to the Liquid Waste Evaporator, where the liquid was heated, 

reduced in volume, and stored in an underground tank. Soil, airborne, and groundwater 

contamination have resulted from these operations. Cleanup activities are under way to reduce 

human health and environmental risks to acceptable levels.

To date, all HLLW has been removed from the tank farm and solidified through calcination. Removing 

the sodium-bearing waste remains one of DOE's highest priorities.
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Internal exposure potential existed from airborne radioactivity created during flushing operations, 

valve repairs, or other maintenance activities in contamination areas from of mixed fission products

and activation products. 

External exposure

2.3.5

occurred during work in the contamination areas and on valve changes and piping 

maintenance where exposure rates measured as high as 500 rad hr-1. 

To remove liquid from the waste, the AEC developed a fluidized-bed calcination process, and built it 

at INTEC. Scientists at Argonne National Laboratory (ANL) tested the method in small-scale models 

in 1955. The process not only solidified the waste, but the product was granular, free flowing, and 

easily handled by pneumatic transport techniques. Phillips engineers started designing the Waste 

Calcining Facility (WCF) in 1956, and construction of WCF-1 started in 1958. The facility was 

constructed just east of the main INTEC process building and south of the storage tanks. Thick 

concrete shielding walls surrounded the process cells, which were below grade. The first campaign 

lasted until October 1964. Two 300,000-gallon tanks and part of a third were emptied before the 

campaign was forced to stop because it had filled all available calcine bins. Half a million gallons of 

liquid had been transformed into 7,500 bulk cubic feet of solid waste. This was a reduction in volume 

of more than 9 to 1. The gases leaving the stack included some 90Sr and 106Ru, but the levels were 

below guideline limits (AEC 1969). 

Waste Calcining Facility-1 and New Waste Calcining Facility, December 1963 To 

2002

In 1982, a new calciner replaced WCF-1. The New Waste Calcining Facility (NWCF) converted liquid 

high-level radioactive waste from the Tank Farm into a granular solid similar in consistency to sand. 

The liquid waste was drawn from underground storage tanks and sprayed into a calciner vessel 

superheated by a mixture of kerosene and oxygen. The liquid evaporated while radioactive solids 

adhered to the granular bed material in the vessel. The off-gases were treated and monitored before 

release to the environment, and the residual calcine solids were transferred to large stainless-steel 

structures encased in thick concrete vaults called bin sets. The calciner was shut down in May 2000 

while DOE evaluated whether to upgrade it to meet new emissions standards or develop a new 

technology to treat the remaining liquid in the Tank Farm. The calciner operated one last run in 2002 

to eliminate the remaining HLLW in the 300,000-gallon storage tanks. HLLW typically contained 300 

Ci m-3. 

The NWCF was the location of a decontamination facility used for cleanup of radiologically 

contaminated materials from INTEC and occasionally from other INEEL processes. The 

decontamination facility continues to function to support INEEL cleanup activities.

Internal exposure potential existed from releases of contamination to the occupational environment 

due to leaks from piping breaks, equipment failures, and other actions that permitted unplanned 

releases and from decontamination activities that would create airborne radioactivity. 

External exposure resulted from routine maintenance on the transfer piping and associated valves 

and equipment. Calcine process cell entries have been made for cleanup and maintenance activities 

with radiation fields measured to 50 rad hr-1 beta/gamma for brief periods. External exposure 

continues to accumulate from the decontamination facility. 
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2.3.6

The INEEL Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) 

Disposal Facility (ICDF) Complex is a new engineered facility south of INTEC and adjacent to the 

existing percolation ponds. This facility was planned to begin operation in July 2003 (INEEL 2003). It 

is designed and authorized to accept wastes from INEEL CERCLA actions. The ICDF Complex 

includes the necessary subsystems and support facilities to provide a complete waste management 

system. The major components include disposal cells (landfill), an evaporation pond (consisting of 

two cells), and the Staging, Storage, Sizing, and Treatment Facility (SSSTF). The Complex covers 

approximately 40 acres, with a landfill disposal capacity of approximately 510,000 yd3

. The 

evaporation pond is designated as a Resource Conservation and Recovery Act (RCRA) Corrective 

Action Management Unit (CAMU) and is the disposal site for leachate and other aqueous wastes 

generated as a result of operating the ICDF Complex. In addition, other aqueous wastes such as 

existing purge water can be disposed in the evaporation pond in accordance with the ICDF 

evaporation pond Waste Acceptance Criteria (WAC).

INEEL Comprehensive Environmental Response Compensation and Liability Act 

Disposal Facility Complex, July 2003 to present

Internal exposure from airborne radioactivity would exist if the integrity of the packaged material is 

compromised during the handing and storage process. 

External exposure 

2.3.7

occurs when shipments of radioactive materials consisting of mixed fission 

products and activation products are placed in storage at the facility. Normal radiation levels are not 

permitted to be greater than 200 mrem hr-1 at any edge of the transporting vehicle. Higher radiation 

fields might be permitted, under special conditions, by proper management authority. 

The Independent Spent Fuel Storage Installation (ISFSI) is a new Nuclear Regulatory Commission 

(NRC)-licensed dry storage area for spent fuel and debris from the TMI accident. Fuel and debris 

were transferred to the INEEL Test Area North for examination, study, and storage after the accident. 

The fuel and debris were transferred to the ISFSI, which provides safe, environmentally secure, 

above-ground storage in metal casks inside concrete vaults. The transfer was completed in mid2002.

TMI-2 Independent Spent Fuel Storage Installation – CPP-1774 to present

Internal exposure potential could exist from leaks or off-gassing from the storage containers.

External exposure 

2.4 ARGONNE NATIONAL LABORATORIES-WEST, February 1951 to present

occurred during the unloading and placement of fuel in the storage vaults.

Argonne National Laboratory - West (ANL-W) was originally known as “the Idaho Division” of ANL. 

ANL-W is an extension of ANL-E near Chicago. ANL is a DOE research laboratory operated by the 

University of Chicago. The original site is in the southwest portion of INEEL, approximately 18 miles 

via Highway 20/26 east of Arco, 40 miles via Highway 26 northwest of Blackfoot, or 50 miles west via 

Highway 20 from Idaho Falls. The original ANL-W location is now a National Historic Landmark and is 

the site of the now decommissioned EBR-I, Boiling Water Reactor (BWR) Experiment, Argonne Fast 

Source Reactor (AFSR), and Zero Power Reactor No 3 (ZPR-III) facilities. 

The present ANL-W site is in the southeast portion of INEEL, about 35 miles west of Idaho Falls. 

There are 52 major buildings at this site, including reactor buildings, laboratories, warehouses, 

technical and administrative support buildings, and craft shops.
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2.4.1

EBR-I, the first reactor built at INEEL, was a NaK-cooled, solid-fuel (enriched uranium), unmoderated 

heterogeneous fast reactor designed for full-power operation at a level of 1 MW. It was built to 

explore the possibilities of breeding nuclear fuel and for the use of liquid metal cooling. A blanket of 

238U around the core provided the fertile material in which nuclear material breeding took place. 

Because the primary coolant was intensely radioactive during and shortly after operation, all primary 

components were enclosed in concrete-shielded cells. The secondary coolant, which was 

nonradioactive, required no shielding (Kittel, Novick, and Buchanan 1957). The facility was entirely 

within a single building of brick, concrete, and steel. Construction on EBR-I began in May 1949 and 

was complete in April 1951. Reactor startup occurred on August 24, 1951.

Experimental Breeder Reactor No. I, April 1951 to December 30, 1963

On November 29, 1955, the reactor suffered a 40 to 50% core meltdown. Radiation detection 

instrumentation measured radioactivity in the building above normal background levels, and all 

personnel were evacuated. After the partial meltdown, the core assembly was removed from the 

reactor using a temporary cave constructed on the reactor top and shipped to ANL- E (Kittel, Novick, 

and Buchanan 1957). The core was replaced and the reactor remained operational until December 

30, 1963. On August 26, 1966, EBR-I was dedicated as a National Historic Landmark. The principal 

radiological activity associated with the coolant during operation and shortly after shutdown was 24Na, 

τ1/2 = 15 hr. The saturation level at full-power operation was approximately 24 μCi gm-1. The second 

most significant activity was about 2 μCi gm-1 137Cs, which apparently entered the system during and 

after the meltdown incident. No other long-lived activity was identified in the primary coolant. Shortlived activity in the form of 133Xe and 135Xe was observed in the cover gas (Haroldson et al. 1963).

Internal exposure potential existed from airborne radioactivity from mixed fission products and 

activation products. 

External exposure

2.4.2

was received by personnel from mixed fission products and activation products 

during activities associated with reactor operation and maintenance.

The Boiling Water Reactor Experiment No. 1 (BORAX-I) was an open-top, water-cooled, watermoderated, boiling-water reactor used to conduct a series of nondestructive experiments in the latter 

part of 1953 and early summer of 1954. The reactor was built in an excavated area of earth 

approximately one-half mile northwest of EBR-I and was housed in a 10-foot-diameter tank open to 

the atmosphere. The control room was approximately one-half mile away near the EBR-I reactor. 

Boiling Water Reactor Experiment No. 1, Late 1953 to July 22, 1954

BORAX I was intentionally destroyed in its final experiment on July 22, 1954. The explosion scattered 

fuel plate fragments and other debris over an area of approximately 200 feet by 350 feet south of the 

reactor area fence. Instrumentation at the control center showed an instantaneous radiation level 

higher than 500 mrem hr-1, which decreased in about 30 seconds to 25 mrem hr-1 and within 5 

minutes to less than 1 mrem hr-1. A detailed discussion of this incident is available (Brodsky and 

Beard 1960).

Internal exposure may have occurred from airborne radioactivity during operation and other activities 

associated with a BWR, the core destruction, coolant, and clean-up activities.

External exposure occurred from direct radiation associated with the reactor operation maintenance 

activities from mixed fission products and activation products. 
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2.4.3 Boiling Water Reactor Experiment No. 2, October 19, 1954, to March 1955

BORAX-II was designed and built to replace BORAX-I to investigate a new reactor that would more 

closely approximate the characteristics of a practical power reactor operating on the boiling-water 

principle. BORAX-II, a short distance northeast of the BORAX-I site, was built in 1954 and became 

operational on October 19, 1954. BORAX-II was significantly larger than BORAX-I. The vessel was 

shielded by concrete and housed in a sheet metal building. Tests of new core combinations were 

tried using varying enrichments of 235U in metal fuel plates. It was a boiling-water system operating at 

300 psi, making it essentially a power experiment. The power level was about 6.4 MW (t) but, 

because it had no turbine generator, it produced no electricity. The energy produced was released in 

the form of steam. In 1955, a turbine generator was added to BORAX-II and testing was done to 

demonstrate that turbine contamination would not be a significant problem in boiling-water reactors. 

Internal exposure was possible from MAP associated with work activities from the reactor coolant. 

External exposure occurred during routine operations and with loading and unloading the reactor. 

2.4.4 Boiling Water Reactor Experiment No. 3, June 9, 1955, to December 1956

As a result of the BORAX-I and BORAX-II tests, a program began in March of 1955 to modify the 

BORAX-II reactor to produce electrical energy. The modified reactor became BORAX- III. The 

previous BORAX reactors were not designed to produce electricity, so a turbine generator was added 

to the facility to convert thermal energy to electricity. The modified facility was capable of generating 

12 MW of thermal energy and 2,300 kW of electrical energy. For 2 hours on July 17, 1955, BORAX-III 

generated approximately 2,000 kW of electricity; 500 kW were used to power the BORAX-III facility, 

1,000 kW were used to power the CFA at the NRTS, and 500 kW were used to light the entire town of 

Arco, Idaho. BORAX-III became the first nuclear power plant in the world to generate electricity for an 

entire city. It was operational from June 9, 1955, until sometime before December 3, 1956. 

Internal exposure was possible from MAP associated with work activities from the reactor coolant.

External exposure occurred from routine operation and loading and unloading of the reactor.

2.4.5 Boiling Water Reactor Experiment No. 4, December 3, 1956, to June 1958

BORAX-IV, the successor to BORAX-III, began operation in December 1956. This reactor, with a 

design power of 20 MW (t), was used principally to test high-thermal-capacity fuel elements made 

from ceramics of uranium and thorium. Like the previous BORAX reactors, BORAX-IV was a boilingwater reactor operating at 300 psig. It was capable of producing 2.5 MW of electricity. It was brought 

to criticality on December 3, 1956, at atmosphere pressure. It operated with a core of uraniumthorium fuel elements until April 17, 1957. Beginning in May 1957, it was operated with a 59-element 

core at 300 psig and 216 oC and continued intermittent operations until December 5, 1957. After the 

core was revised to increase the maximum power, the reactor was restarted on February 19, 1958, to 

evaluate the effect of operating with a fuel element defect and to locate defective elements in the 

core. BORAX-IV released approximately 4,565 curies of short-lived radionuclides to the atmosphere 

in March 1958 (Novick 1958). It operated until June 1958. The following measurements were made 

during reactor operations:

1. Radiation levels of the steam plant equipment
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2. Quantitative determination of fission gases 138Xe and 88Kr, which were released to the 

atmosphere through the air ejector

3. Analysis of reactor water, condensed steam before the turbine, and condensed steam after the 

turbine (hot-well condensate) for fission products

4. Area contamination downwind from the reactor

Internal exposure may have occurred during work with the defective fuel elements or during planned 

releases of short-lived fission activity and from airborne MAP/MFP associated with the reactor 

coolant. 

External exposure occurred during operation and work with loading and unloading the reactor. 

2.4.6 Boiling Water Reactor Experiment No. 5, February 9, 1962 to September 1964

BORAX-V was a flexible boiling-water reactor with the same configuration as BORAX-IV, used 

primarily for testing nuclear superheating concepts. The facility was operational from February 9, 

1962 until September 1964. 

Internal exposure may have occurred from coolant and airborne activity during routine BWR operation 

with fuel elements made from ceramics of uranium and thorium, and associated maintenance work. 

External exposure occurred from routine activities associated with reactor operation and maintenance. 

2.4.7 Zero Power Reactor No. 3, October 1955 to November 1970

ZPR III was a low-power, split-table reactor that achieved criticality by bringing two halves of a fuel 

configuration together. It was used to determine the accuracy of predicted mass geometries and 

critical measurements for fast reactor core designs.

The building consisted of a reinforced concrete high bay assembly room and a one-story section 

containing the control room, work room, vault, laboratory rooms, offices, etc. The assembly (reactor) 

room of reinforced concrete was approximately 45 ft x 42 ft x 29 ft high (Brittan et al. 1961).

The assembly machine was a platform on which two tables or carriages were mounted, one of which 

was moveable. Half of the reactor was built up on each carriage by inserting drawers containing the 

reactor material into a matrix structure. Each half of the assembly contained five safety control rods 

and a 15-curie polonium-beryllium neutron source. A hinged platform could be swung into place 

between the halves to stand on while loading or unloading the machine. 

The storage vault room was approximately 29 feet long by 26 feet wide with walls and roof of 

reinforced concrete 9 inches thick. The fuel slugs were stored either on racks or in special “birdcage”-

type containers that limited the storage density to 2 kilograms of 235U per cubic foot.

Internal exposure potential existed from possible airborne radioactivity during maintenance operations 

from MFP/MAP. 

External exposure occurred maintenance activities and during loading and unloading of the reactor 

fuel. 
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2.4.8 Argonne Fast Source Reactor, October 29, 1959, to Late 1970

AFSR was a small fast reactor facility designed to produce neutrons for the development of special 

equipment for the fast reactor programs of EBR-I, EBR-II, and ZPR-III. The reactor, with a design 

power of 1 kilowatt, was in a prefabricated Butler-type building built in 1958 near the EBR-I facility with 

its own heating and air compressor plant. No water was plumbed into the building. Control and 

safety mechanisms were in a pit below the reactor. The reactor, designed to supply both fast and 

thermal neutron fluxes for laboratory experiment, was built around a cylindrical core of solid highly 

enriched uranium with a blanket of solid DU (Brunson 1959). Reactor startup occurred on October 

29, 1959; the reactor was operational until sometime in the late 1970s when it was moved to a new 

location adjacent to the ZPPR facility at the ANL-W site. The reactor is now shut down and defueled.

Internal exposure may have occurred during routine operations that could create airborne 

radioactivity. 

External exposure occurred from maintenance activities and loading and unloading fuel.

2.4.9 Transient Reactor Test Facility, February 23, 1958, to April 1994

TREAT was an air-cooled thermal heterogeneous system designed to evaluate reactor fuels and 

other material under conditions simulating various types of reactor excursions. Construction began in 

February 1958 and ended in November 1958, and criticality was achieved on February 23, 1959. The 

TREAT complex comprises a reactor building and a control building approximately 1 mile northwest of 

the EBR-II containment building (Freund et al. 1960).

The reactor building features a high bay section and an adjacent service wing. The high bay section 

contains the reactor, fuel storage pit, instrument room, and the basement subreactor and equipment 

rooms. The control building is a single-story concrete block structure approximately one-half mile 

from the reactor that contains control panels and necessary instrumentation for remote control of the 

reactor.

Shielding permitted personnel access around and atop the reactor during steady-state operations at 

100 kW. Access to the subreactor room is controlled during steady-state operation. Prior to transient 

operations, the building was evacuated of all personnel. General neutron and gamma radiation levels 

at a distance of 10 feet from the reactor during operations at 100 kW were (Freund et al. 1960):

• Fast Neutron Negligible

• Thermal Neutrons 50-1,500 n cm-2 sec-1

• Gamma 5-8 mrem hr-1

Internal exposure may have occurred during routine operations that could create airborne 

radioactivity; however, it was not expected to occur.

External exposure occurred from routine operations. 

2.4.10 Experimental Breeder Reactor No. II, May 1961 to September 30, 1994

EBR-II, at the ANL-W site, is a liquid sodium-cooled, unmoderated, heterogeneous fast breeder 

reactor rated at 62.5 MW thermal, with an intermediate closed loop of secondary sodium and a steam 

plant capable of producing electrical power through a conventional turbine generator. A fuel 

processing facility is attached to the reactor. EBR-II was designed to prove the breeding of fuels, the 
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feasibility of a central power station, and onsite fuel processing. These objectives were met in the late 

1960s, and the role of EBR-II changed to that of a test reactor. Construction of EBR-II ended in May 

1961, and the reactor reached criticality on September 30, 1961. It operated until September 30, 

1994, when it was taken to a subcritical configuration and shut down to start a defueling operation. 

On January 19, 2001, ANL-W verified that the liquid metal sodium coolant had been completely 

drained from the reactor vessel. At present, the reactor is defueled, the sodium systems have been 

drained, and the power plant is depressurized.

The reactor was submerged in a primary tank containing approximately 90,000 gallons of liquid 

sodium. This tank was suspended in an airtight steel-shell containment building of 1-inch-thick steel 

plate, which would contain an accidental release of fission products, etc., from the primary system. 

The structure of the primary system is designed to contain the energy release associated with a 

reactor incident. The reactor building is designed to confine the effects of a maximum sodium-air 

interaction caused by a major sodium release. The reactor consists of an enriched core surrounded 

on all sides by a fertile blanket of depleted uranium (McVean et al. 1962; Koch et al. 1957)

The Sodium Plant contains the pumping, purification, and storage facilities for the secondary sodium 

system. It also contains a receiving station for the sodium. The building was not normally occupied. 

The primary and secondary coolant from EBR-II is converted in the Sodium Processing Facility (SPF) 

from its elemental, chemically unstable form, to a chemically stable composition suitable for landfill 

disposal.

The Fuel Manufacturing Facility (FMF) is a secure facility designed for the fabrication of EBR-II fuel. 

The FMF vault stores special nuclear material in support of the EBR-II shutdown.

An additional building, the Laboratory and Office Building near the EBR-II plant, provided supporting 

analytical and personnel facilities

2.4.11 Hot Fuel Examination Facility, 1964 to 1968

The Hot Fuel Examination Facility (HFEF) complex comprises two hot cell facilities, HFEF/South and 

HFEF/North. HFEF/South. Originally known as the Fuel Cycle Facility and/or the Fuel Conditioning 

Facility, was used to demonstrate remote processing and refabrication of uranium-fission metal-alloy 

driver fuel elements in a closed cycle with EBR-II. To accomplish the process functions, the following 

major components were necessary:

1. The passageway between the reactor building and the operating area at the Air Cell

2. The Air Cell

3. The Argon Cell, with a subcell for equipment

4. The operating area surrounding the cells, including auxiliary laboratories

5. The service area (beneath the operating area)

6. Suspect stack, including ventilation systems and fuel off-gas holdup tanks

7. Center Control Area

8. Analytical laboratory and service building
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Some 35,000 fuel elements were remotely reprocessed and refabricated into EBR-II subassemblies 

between 1964 and 1968. HFEF/South contains two large heavily shielded hot cells; one with an inert 

gas (argon) atmosphere, the other with an air atmosphere. The shielding walls of both cells are of 

high-density concrete. The HFEF/South air cell was decontaminated and refurbished in 1969 and 

again in 1976. The argon cell was deactivated in 1977. In June 1968, personnel entered the argon 

cell for the first time since it was activated in 1964.

HFEF/North is a large alpha-gamma hot cell facility that was activated in March 1975. This facility 

provided the capability for post irradiation and nondestructive or destructive examination of fuel and 

material experiments irradiated in EBR-II. HFEF/North contains two hot cells, one with an argon gas 

atmosphere and the other with an air atmosphere. The air atmosphere cell was known as the 

decontamination cell. The shielding walls of both cells are of high-density concrete (Baca 1979).

Remote characterization of material to be shipped to the Waste Isolation Pilot Plant (WIPP), in New 

Mexico, for disposal takes place in the Waste Characterization Area (WCA) of the HFEF high bay.

Internal exposures might have occurred during cell entries when suspended radioactive contamination 

materials could cause airborne radioactivity from mixed fission products and activation products. 

External exposure occurred when entries to the hot cell were made after experiment processes or 

during equipment maintenance and refurbishment. 

2.4.12 Zero Power Physics (Plutonium) Reactor, April 18, 1969 to April 1992 (standby)

ZPPR is a split-table critical facility approximately 300 meters from EBR-II in the ANL-W area and 

about 3 miles north of U.S. Highway 20. The facility is divided into two areas, the mound area and the 

support wing. The mound area consists of the reactor cell, fuel storage vault, workroom, and 

equipment rooms as well as access and escape tunnels. The reactor cell is a 50-foot-diameter 

circular room with floor and walls of reinforced concrete. The roof is composed of layers of washed 

and dried sand and gravel supported by a catenary cable network.

The basic element of the ZPPR is a bed-and-table system, which holds the matrix assembly. The two 

tables, one moveable and one stationary, are supported on a cast steel bed. The control and safety 

rod drives are mounted near the rear of each table. The main floor consists of the reactor control 

room, offices, an electronics shop, and a core coating room. The core coating room, adjacent to the 

control room and the entrance to the mound area, is used to clean core stimulants such as 235U and 

stainless steel. The room contains two hoods for handling suspect materials and a core coating 

machine that is used primarily to dry and coat depleted uranium with a protective film.

Internal exposure potential was minimal due to the use of hoods and other protective equipment. 

External exposure occurred from working with reactor processes, loading and unloading fuel, etc. 

2.4.13 Neutron Radiography Facility, October 1, 1977 to present

The Neutron Radiography (NRAD) Facility is a 250-kW, steady-state training research and isotope 

(built by General Atomics) (TRIGA) reactor in the basement beneath the HFEF/North main cell. The 

reactor core consists of fuel elements surrounded by graphite assemblies. The core is submerged in 

a water-filled tank. NRAD began operation on October 12, 1977, with two radiography stations. The 

East station services the hot cell complex where specimens can be radiographed without removing 

them from the hot cell environment. The North station is outside the cell in a separate, clean shielded 
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location for the radiography of irradiated or unirradiated items without introducing them into the 

contaminated cell. Cask handling and specimen shielding allow for full-size reactor assemblies. The 

radiography room is easily accessible for development work (Richards and McClellan 1979).

NRAD has limited irradiation capabilities in the core. It has a water-filled port at the center of the core 

and a dry port at the edge of the core. NRAD operates a MF Physics linear particle accelerator that is 

used for nondestructive assays of waste and expended nuclear fuel.

Internal exposure potential exists from possible airborne radioactivity primarily from the hot cell 

environment from mixed fission products and activation products. 

External exposure occurred during sample handling and maintenance associated with radioactive 

samples. Remote handling techniques are used to minimize dose. 

2.4.14 Fuel Assembly and Storage Building, 1970 to present

The Fuel Assembly and Storage Building (FASB) is a multipurpose facility that supports development 

of low-enrichment uranium fuel for research reactors, storage of spent fuel, and examination of the 

condition of other experimental projects. The East (clean) room houses offices, restrooms etc. The 

West room contains a vault for the storage of nuclear material. It also contains equipment for 

performing materials testing and for preparing metallurgical samples, and inert atmosphere glove 

boxes and hoods. The facility no longer does fuel assembly, but other radiological work is ongoing.

Internal exposure may occur from airborne radioactivity associated with the described processes from 

the uranium fuel and spent fuel examinations. 

External exposure occurred from movement of radiological samples and reactor fuel. 

2.4.15 Other ANL-West Support Facilities 

At ANL-W, an Analytical Laboratory (AL) provides the capability for performing chemical and physical 

measurements of both radioactive and non-radioactive samples. The facility includes six analytical 

hot cells and both general and specialized chemistry laboratories. Personnel were subjected to 

radiation levels above building background on occasion in the Junior Cave Area. 

The Radioactive Scrap and Waste Facility (RSWF) provides in-ground retrievable dry storage for 

nuclear fuels and other highly radioactive scrap and waste, and interim storage for EBR-II spent fuel.

The Radioactive Liquid Waste Treatment Facility (RLWTF) evaporates low-level radioactive liquid 

waste generated at ANL-W facilities into solidified residue packaged in shielded containers

The Industrial Waste Pond (IWP) is an unlined evaporative seepage pond that is fed by a system of 

drainage ditches. It has been used since 1964 to receive wastewater from a number of sources. The 

largest sources of liquid industrial waste going to the IWP are blowdown effluents from the main and 

auxiliary cooling towers, auxiliary boilers blowdown; water from once-through air conditioning, and 

cooling water from other sources. There might be inadvertent low-level radioactive contamination in 

this pond.

The three Sanitary Sewage Treatment Ponds (STP) north of EBR-II cover an area of about 2 acres. 

These ponds are not suspected to be radiologically contaminated.
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Internal exposure is possible from airborne radioactivity associated with the various types of samples 

worked in these facilities. 

External exposure occurred from working with the variety of radioactive material associated with these 

facilities (mixed fission products, activation products and transuranics). 

2.5 RADIOACTIVE WASTE MANAGEMENT COMPLEX, MAY 1952 to present

The RWMC is 51 miles west of Idaho Falls. The first 13 acres were fenced off in May 1952 for 

shallow-land disposal of solid low-level radioactive waste, and burial of transuranic waste and 

hazardous substances such as organic and inorganic chemicals. The current RWMC mission 

includes interim storage of transuranic waste and shipment of stored transuranic waste to the WIPP 

near Carlsbad, New Mexico, for permanent disposal.

For the first 2 years, only low-level radioactive waste was buried at the RWMC. In 1954, Rocky Flats 

began shipping defense waste with transuranic elements. By 1957, the original 13 acres were nearly 

filled, and the RWMC was expanded to 97 acres. In 1970, it was expanded again to 168 acres. After 

1970, transuranic waste was placed in retrievable storage on asphalt pads and covered with an 

earthen berm. This waste, in drums and boxes, is stored in engineered modules. From 1970 to the 

present, low-level waste has been disposed of in 20 pits, 58 trenches, and 21 soil vault rows. INEEL 

has been repackaging (as needed) and shipping transuranic waste to the WIPP.

Early packaging configurations were thought not suitable for extended storage and could present 

future hazards to the workers, the public, or the environment. Early wastes were pushed out of trucks 

into open pits or trenches and covered with soil by heavy equipment, which might have occasionally 

damaged containers in the covering process. Long-tongued dump trailers were used to minimize 

exposure to personnel from the waste. Other types of heavy equipment such as cranes and cherry 

pickers were used to pick shielded containers from trucks for unloading the contents into soil vaults, 

etc. There were occasional brief exposures greater than 100 mrem hr-1 (photon energy ≥ 250 keV).

In addition to administrative buildings, the RWMC has the following facilities and processes:

2.5.1 Subsurface Disposal Area 

The 97-acre Subsurface Disposal Area (SDA) is in the western section of the RWMC. It contains an 

active shallow-land-burial area for the permanent disposal of solid low-level waste. It also contains 

pits and trenches where mixed transuranic and low-level waste was buried between 1954 and 1970. 

Solid waste from the Rocky Flats Plant in Colorado comprised a large fraction of the waste received at 

the RWMC. For example, in 1969 approximately 250,000 cubic feet of waste from the Rocky Flats 

Plant with a reported activity of more than 35,000 curies was buried at the RWMC. Rocky Flats waste 

was usually contaminated with plutonium isotopes and 241Am (RAC 2002, pp. 26-27).

2.5.2 Intermediate Level Transuranic Storage Facility

The Intermediate Level Transuranic Storage Facility (ILTSF) has had 53 drums of 233U stored in metal 

cargo containers in an open yard surrounded by concrete block shielding.
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2.5.3 Transuranic Storage Area

The 56-acre Transuranic Storage Area (TSA) is in the southern section of the RWMC dedicated to 

storage of contact-and remote-handled packages of solid transuranic waste. This waste was received 

at INEEL after 1970 and was stored above ground.

2.5.4 Stored Waste Examination Pilot Plant

The Stored Waste Examination Pilot Plant (SWEPP) certifies waste to ensure that it meets repository 

acceptance criteria. Examinations used a shielded 450-kVp X-ray facility and a neutron assay system 

that developed 14-MeV neutrons. Certified waste has been stored temporarily in permitted storage 

areas until it is packaged and sent to the WIPP. Noncertified waste is segregated until processing is 

available that will enable it to meet acceptance criteria. 

2.5.5 Transuranic Package Transporter Loading Station

This loading station is used to load transuranic waste into Transuranic Package Transporter 

(TRUPACT)-II shipping containers for shipment to the WIPP. Responsibility for the station has been 

turned over to British Nuclear Fuels Limited (BNFL), Inc.

2.5.6 Advanced Mixed Waste Treatment Project 

BNFL operates the Advanced Mixed Waste Treatment Project (AMWTP) under contract with DOE. 

The facility, which is under construction, will retrieve and process approximately 65,000 cubic meters 

of mixed transuranic waste in temporary storage at the TSA, treat the waste to meet environmental 

laws and disposal criteria, and package it for shipment to the WIPP.

Exposures below represent all of the subsections of 2.5. 

Internal exposure may have resulted when workers were proximate to breached and or leaking waste 

containers containing mixed fission products, transuranics, activation products, etc. from the multiple 

materials in storage at the RWMC. 

External exposure might have resulted from working adjacent to the various waste containers in 

storage or disposal in the form of intermediate-level, low-level, transuranic, and mixed waste, and 

from SWEPP X-ray (30-450 keV) and neutron (2-20 MeV) waste package examination and 

certification processes. Approximately 150 6M drums containing as much as 500 grams of 232/233U per 

drum have been stored under earthen covers on pads in the TSA (232U is always present as a 

contaminant with 233U and accounts for a significant in-growth of high-energy gamma emitters). 

Depending on the waste involved, radiation exposure levels might have been near background levels, 

or could have exceeded a few hundred mrem hr-1 with photon energies greater than 250 keV for short 

periods. 

Radionuclides at the RWMC cross the spectrum based on the character of operations at this facility. 

Those listed in Table 2-3 would represent many of the long-lived fission products. Activation products 

from reactor facilities were also a concern for external exposure.

2.6 CENTRAL FACILITIES AREA, AUGUST 2, 1943, to present 

The CFA is the main service and support center for INEEL programs and the other technical areas on 

the 890-square-mile Site. It is about 50 miles west of Idaho Falls, Idaho, just north of Highway 20 in 
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the south-central portion of the Site. Most CFA buildings and activities support transportation, 

maintenance, capital construction, environmental and radiological monitoring, security, fire protection, 

warehouses, calibration laboratories, and a cafeteria. There is a small amount of R&D work. What is 

now the INEEL started as an offshoot of the Naval Proving Ground command area (dedicated August 

2, 1943) where the Navy tested ordnance from fighting ships. Buildings constructed by the Navy 

became the staging area for INEEL development that began in earnest in1950. The area continued to 

expand as a central service area for the NRTS. Eventually it acquired the name “Central,” or more 

officially, CFA. 

In the late 1950s and into the 1960s, small amounts of radioactivity were processed through a 

Sewage Treatment Plant, CF-674, to a drying pond. Most of the radioactivity was from the “hot” 

laundry, although small amounts could enter from CF-656 and CF-690. CF-656 was a Reactor 

Engineering Laboratory where tracer-level radionuclide and chemistry work occurred. CF-690 

included the laboratory where analytical chemistry was done on bioassay samples, naturally occurring 

radionuclides, and other special projects. The dispensary included an X-ray unit for medical use (30-

250 keV). The exposures listed below are for all of the subsections of section 2.6.

Internal exposure. CFA internal exposure potential was primarily from mixed fission products, 

activation products and transuranics associated with articles cleaned at the laundry facilities. Laundry 

facility workers were included in bioassay programs.

External exposure. CFA external exposure potential is greatest from calibration sources and X-ray 

equipment at the HPIL and the DOELAP Irradiation Facility with photon energies greater than 250 keV 

from calibration sources and 30-250 keV for X-ray photons. Neutron energies range from 2 to 20 

MeV from the 252Cf source and the AmBe source. Personnel working within radiological areas must 

wear dosimetry devices. 

2.6.1 The Hot Laundry 

The laundry, in the east portion of CFA, washed coveralls and other protective clothing items used for 

radiological work. The laundry drain went to a septic tank and drain field with other sanitary waste. 

The laundry facility and drain field(s) are sources for low-level radioactive contamination, which covers 

the spectrum inherent to work in radiological contamination areas. Two laundry facilities next to one 

another (old and new facility) serviced INEEL. They are no longer operating.

2.6.2 The Health Physics Instrument Laboratory

The Health Physics Instrument Laboratory (HPIL), CFA-633, was a calibration facility used for 

radiological instrumentation standardization. The HPIL used 252Cf neutron sources, and alpha, beta, 

and gamma sources for health physics instrument calibrations. All of the sources were sealed. A 

new facility, CF-1618, completed in late 2002, includes six automated irradiator systems and provides 

expanded neutron, gamma, and X-ray irradiation capabilities. The higher level sources require 

external exposure control, and personnel in the radiological work area must wear applicable 

dosimetry.

2.6.3 DOE Laboratory Accreditation Procedure Irradiation Facility

The DOELAP Irradiation Facility, CF-636, is an above-ground shielded bunker that houses radioactive 

sources used by DOE-ID for radiation measurement equipment calibrations. The bunker is on the 

access road east of the main road into the CFA. It houses an X-ray facility, seven 1-Ci 241Am sources, 
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beta sources, and two 137Cs sources (20 and 1,000 Ci). At one time there was an AmBe neutron 

source at this facility.

2.6.4 Radiological Environmental Science Laboratory

The RESL facility, operated by DOE-ID in CF-690, evaluates low-level environmental and other 

laboratory samples that pose minimal radiological risk from internal or external pathways. The 

Dosimetry facility, also in CF-690, has been used to process external dosimetry devices such as film, 

TLDs, etc. 

2.6.5 CF-674 Building

CF-674 was used from 1953 to 1969 as a Chemical Engineering Laboratory (CEL) to conduct calcine 

experiments on simulated nuclear waste. The experiments created liquid waste streams discharged 

to a pond designated as CFA-04. This waste stream was contaminated with calcine that contained 

low-level radioactive waste (DOE-ID 2003).

2.7 TEST REACTOR AREA, MARCH 31, 1952, TO PRESENT

The TRA is approximately 5 miles north of Central Facilities. Eight reactors have been built and 

operated in the TRA. Three of the reactors – Materials Test Reactor, Engineering Test Reactor 

(ETR), and Advanced Test Reactor (ATR) – were high-flux reactors designed for materials testing. 

The remaining five – Reactivity Measurement Facility (RMF), Advanced Reactivity Measurements 

Facility No. 1 (ARMF-1), ARMF-2, Engineering Test Reactor Critical (ETRC), and Advanced Test 

Reactor Critical (ATRC) – were low-power reactors designed to perform reactivity measurements. At 

present, only the ATR and ATRC are operational.

Other TRA facilities of radiological concern are the TRA Hot Cells, Gamma Facility, Radiation 

Measurements Laboratory, Radiological Chemistry Laboratory, Liquid Waste Disposal Ponds, and 

High-Level Liquid Waste Disposal Tanks and Transfer Facility.

All personnel entering the TRA must wear a dosimetry badge, and those who work in or near 

radiological control areas must wear PICs. Personnel working in radiological control areas are on a 

routine bioassay program and receive routine whole-body counts.

2.7.1 Materials Test Reactor, March 31, 1952, to April 23, 1970

The MTR (TRA 603) was the original reactor at the Test Reactor Area and the second reactor to be 

operated at INEEL. Fueled with enriched uranium fuel, water-cooled and water-moderated, this 

reactor was a key part of the AEC postwar reactor development program. It operated at a power level 

of 30 MW until September 1955, when thermal output was increased to 40 MW. It supplied a high 

neutron flux in support of a reactor development program subjecting potential reactor fuels and 

structural materials to irradiation.

The MTR gave the researcher several options to achieve sample irradiation. Lead experiments 

entered from the top of the reactor with positions around the core. Pneumatic ports on the reactor top 

enabled the insertion of capsules for irradiation in the graphite region around the core. A hydraulic 

rabbit system underneath the reactor enabled the insertion of specimens and their discharge to the 

canal during reactor operation. In addition, horizontal and angular beam holes made it possible to 

perform cross-section measurements and other physics research, including several neutron 

experiments. The high-flux radiation fields available in this reactor made it possible to accelerate the 
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screening of test materials. In its early years, the MTR contributed to the design of pressurized-water, 

organic-moderated, liquid-metal-cooled, and other reactors. Its successful operation resulted in a 

family of plate-type reactor fuels.

The MTR logged more than 125,000 operating hours and more than 19,000 neutron irradiations. 

During August 1958, it became the first reactor to operate using 239Pu as fuel at power levels as high 

as 30 MW. In early 1970, the MTR was once again fueled with 239Pu. The last core was named 

"Phoenix" after the legendary bird that lived 500 years, burned itself to ashes, then rose to live again. 

The plutonium cores demonstrated that a plutonium-fueled, water-moderated reactor could be 

controlled satisfactorily. In August 1970, the MTR was again brought to power for a 24-hour run to 

irradiate 1,000 biological samples for iodine analysis.

Internal exposure was most probable during the first few hours of shutdown. When the reactor top 

was removed, airborne fission products would be released. During shutdown, airborne radioactivity of 

mixed fission products and activation products from maintenance activities resulted in some potential 

for internal exposure. 

External exposure occurred during the variety of operations associated with sampling a test reactor 

and the associated maintenance activities. The major contributors to external exposure were mixed 

fission and activation products that emitted beta and gamma radiation with energies typically above 

250 keV. 

2.7.2 Engineering Test Reactor, September 19, 1957, to December 1981 

When the 175-MW ETR started in 1957, it was the largest and most advanced nuclear materials test 

reactor in the world. It provided larger test spaces than the older MTR and a more intense neutron 

flux. ETR fuel, coolant, and moderator materials were evaluated under environments similar to those 

of power reactors. Several experiment loop facilities were designed to test the fuels for the ANP 

Program and the Navy fuel development program.

In 1972, a Sodium Loop Safety Facility was added to the ETR reactor core. With this, the reactor 

played a new role supporting the DOE breeder reactor safety program. ETR test programs were 

related to the core design and operation of breeder reactors. As testing progressed, the reactor was 

modified with a new top closure to accommodate the irradiation loop. Other additions included a 

helium coolant system and sodium-handling system. The ETR was the first complete reactor facility 

to be deactivated and the D&D documented immediately after shutdown.

Internal exposure from airborne radionuclides was minimal during normal reactor operation. 

Exposures may have occurred during shutdown as airborne fission products were often released 

when the reactor top was removed and access was made to the reactor subpile room and experiment 

cubicles for maintenance activities. 

External exposure was received by workers in the reactor area during shutdown and changes of loop 

and lead experiment samples as required. There were cases of significant gamma fields exceeding 

50 mrem hr-1 from fission and activation products in the reactor subpile room, loop cubicles, and 

nozzle trench.

2.7.3 Advanced Test Reactor, July 7, 1967, to present

The ATR is the latest materials testing reactor to be built in the TRA. It simulates the environment in a 

power reactor to study the effect of radiation on steel, zirconium, and other materials. The ATR 
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produces an extremely high neutron flux, making it ideal for materials testing. Target materials are 

exposed to the neutron flux to test their durability in an environment of high temperature, high 

pressure, and high gamma radiation fields. Data that normally would require years to gather from 

ordinary reactors can be obtained in weeks or months from the ATR.

The ATR can operate up to a power level of 250 MW. Its unique four-lobed design can deliver a wide 

range of power levels to nine main test spaces, or loops. Each loop has its own distinct environment 

apart from that of the main reactor core. Smaller test spaces surrounding the loops enable additional 

tests. In addition, ATR produces radioisotopes for use in medicine, industry, and for other research.

Internal exposure might have occurred during reactor shutdown from mixed fission and activation 

products released to the air in the occupied environment of the building. The most probable locations 

would be from the reactor top, experiment cubicles, primary coolant rooms, and subpile room. During 

reactor operation, areas with airborne radiation are exclusion areas.

External exposures are received primarily during reactor shutdown from fission and activation 

products in the fuel, experiments, and associated hardware. Work on the primary system, in the 

reactor tank, or loop cubicles would have associated external exposure. Handling of isotope 

production samples would produce some exposure depending on the nuclide and quantity. Typical 

nuclides generated would include 192Ir, 99Tc, 60Co, and 131I. 

2.7.4 Reactivity Measurement Facility, February 11, 1954, to April 10, 1962

The RMF was a very-low-power reactor in the east end of the MTR canal that operated at a power 

level of 100 or 200 watts. Water was its moderator, reflector, and shield. It was designed to measure 

reactivity changes in materials irradiated in the MTR or ETR. The RMF was used to assay new and 

spent fuel elements and to assist in experiment scheduling by evaluating reactivity losses and flux 

depression caused by in-pile apparatus (Stacy 2000). 

Internal exposure. There was a potential for internal exposure from mixed fission or activation 

products from airborne radioactivity during maintenance activities. 

External exposures were minimal due to the low-power operating level and the depth of the pool. Any 

external exposures would have come from fission and activation products, primarily during fuel 

handling.

2.7.5 Advanced Reactivity Measurement Facility No.1, October 10, 1960, to 1974 

The ARMF-I reactor was in a small pool in the TRA-660 building east of the MTR building. It was 

used to determine nuclear characteristics of reactor fuels and other materials for testing in the MTR. 

2.7.6 Advanced Reactivity Measurement Facility No. 2, December 14, 1962 to 1968

The ARMF-II reactor was in the opposite end of the tank occupied by ARMF-I. It had a “readout” 

system that automatically recorded measurements on International Business Machines (IBM) data 

cards. This refinement over the ARMF-I reactor meant that operators could process data quickly in 

electronic computers. 

Internal exposure. ARMF-I and ARMF-II airborne activity of fission products could exist if a fuel 

element or sample was damaged during handling. 
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External exposure from ARMF-I and ARMF-II was minimal due to their low operating power levels and 

the depth of the canal in which they are located. Exposures would be from fission products and or 

activation products during fuel or experiment handling.

2.7.7 Coupled Fast Reactivity Measurement Facility, 1968 to 1991

When the ARMF-II reactor was modified in 1968, it received a new name, the Coupled Fast Reactivity 

Measurement Facility (CFRMF). The core was modified to produce a region of high-energy neutron 

flux to provide physics information about the behavior of fast (unmoderated) neutrons. Physicists 

studied differential cross-sections and tested calculational methods. The CFRMF contributed to the 

development of fast neutron reactors (Stacy 2000).

Internal exposure potential existed from airborne fission products that may have been released from 

the fuel.

External exposure. Minimal exposure resulted from working on the reactor top to move fuel or 

experimental components. Mixed fission or activation products would be the major contributors.

2.7.8 Engineering Test Reactor Critical Facility, May 20, 1957, to 1982

ETRC was a full-scale, low-power nuclear facsimile of the ETR in TRA-635, similar in function to the 

ARMF and ATRC. It was used to determine the nuclear characteristics of fuel and experiments 

planned for irradiation in ETR and/or the power distribution effects for a given ETR fuel and 

experiment loading. The ETRC enabled operators to predict the nuclear environment when 

completed experiments were removed or new ones added to calculate the experiment irradiation and 

determine core life, control rod withdrawal sequences, reactivity worth, and core safety requirements.

Proposed fuel and experiment loadings were mocked up in ETRC and manipulated until a desired 

power distribution throughout the core was attained, satisfying pertinent safety requirements. ETRC 

low-power tests enabled the ETR to operate without interruption, saving time and money (Stacy 

2000).

Internal exposure from airborne radioactivity of mixed fission products from damaged fuel or leaky 

experimental samples containing activation products is possible. 

External exposure. Irradiated fuel or samples were transferred in or out of the ETRC under water to 

minimize exposure. The operating console was not on the reactor top, which also minimized external 

exposure. External exposure would have been from mixed fission products and activation products 

during loading or unloading fuel or test samples.

2.7.9 Advanced Test Reactor Critical Facility, May 19, 1964 to present

The ATRC performs functions for the ATR similar to those of the ARMF reactors in relation to the 

MTR. It verified for reactor designers the effectiveness of control mechanisms and for physicists 

predictions of power distribution in the large core of the ATR. Low-power testing in the ATRC 

conserved time so the large ATR could irradiate experiments at high power levels, and verified the 

safety of a proposed experiment before it was placed in the ATR (Stacy 2000).

Internal exposure was possible from fission products released from damaged fuel or experiment 

samples containing activation products The damage could occur during transfer in or out of the 

ATRC. Some fuel or samples had been preirradiated in the ATR or other facilities. 
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External exposure occurred during canal work. The operators must on or leaning over the canal 

parapet to work with irradiated fuel or experiment samples being transferred in or out of the core 

under water. During reactor operation, access to the canal parapet is restricted. The ATRC canal is 

21 feet deep. 

2.7.10 TRA Hot Cell Facility 1954 to present

The TRA Hot Cells are southwest of the ETR reactor building. They consist of three separate cells 

with a common operating corridor. The operators are protected by thick concrete walls and special 

viewing windows, behind which they can handle, photograph, mill, measure, and weigh radioactive 

samples. The work in the cells has involved all types of samples including irradiated fuel, 

transuranics, and isotope production material. Irradiated samples, including failed reactor fuel, can be 

put in cell 1 or 3 from a shielded cask outside the building. Each cell has to be entered periodically to 

repair equipment or set up for a new job. Entry is through a shielded door in the back of the cell. 

Internal exposure might occur during cell entries from the samples and the work performed with 

irradiated samples, dust, or particles that could become airborne. 

External exposure occurs from samples that go in the hot cells. Exposure to beta, gamma, and 

neutrons has occurred. Fuel samples of different types and different ages result in mixed fission and 

activation products and transuranics, including 252Cf. Many samples have been prepared as isotopic 

sources for industrial or medical applications. Cell entry and sample handling result in most of the 

external exposure, and some has occurred from sample ports and manipulator removal operations. 

2.7.11 TRA Gamma Facility, 1955 to unknown 

The TRA gamma facility was constructed south of the original TRA main security gatehouse. The 

facility consisted of a 16-foot-deep canal with cadmium buckets designed to hold spent MTR fuel 

elements. Experimental samples were inserted in sample tubes and lowered into extremely high 

gamma fields.

Sponsors have provided a large variety of materials and samples for gamma irradiation, including 

food products and some natural substances such as gold, diamonds, and oil. Irradiated samples, 

which were not radioactive, were surveyed thoroughly for external contamination on removal.

Internal exposure is possible from a spent fuel element used for irradiations which is unlikely to be 

damaged. 

External exposure was minimal due to handling procedures and shielding from the canal and 

transport devices. Fuel was transported in large casks in and out of the canal.

2.7.12 Radiation Measurements Laboratory, 1952 to present

The Radiation Measurements Laboratory (RML), in the MTR west wing, was previously called the 

MTR Counting Room. The RML specializes in measuring quantity and quality of alpha, beta, gamma, 

and neutron radiation samples. A variety of counting equipment and spectrometers are available in 

the RML. Over the years, an endless variety of samples has been brought in for counting. Some of 

the detector shielding was made out of pre-World War II battleship steel. The original equipment has 

been replaced with more modern equipment. 
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Internal exposure is possible from airborne radioactivity from mishandled samples brought in for 

counting. The isotopes encountered would include fission products, activation products, transuranics, 

and more.

External exposures are very low due to the small samples required for the counting equipment. 

Counting room personnel can encounter neutrons and beta and gamma emitters. 

2.7.13 Radiochemistry Laboratory 

The Radiochemistry Laboratories are in the MTR west wing. They are used to support the RML and 

to perform independent R&D work. Investigators study ways to produce and purify medical 

radioisotopes and the effects of radiation on hazardous waste.

Laboratories 109 through 112 were used primarily for chemistry analysis of reactor primary systems 

and loop experimental coolants. The predominant radioactivity was mixed fission and activation 

products. The south extension to the MTR Wing is the Alpha Laboratories, designed for the safe 

handling of hazardous alpha emitters such as 233U, 239Pu, 241Am, and other transuranics, including 

252Cf. 

Internal exposure potential existed from airborne radioactivity from the large variety of samples. 

External exposures occur when irradiated samples are brought into the laboratories. One laboratory 

has a shielded box, similar to a small hot cell, used for handling highly radioactive samples. 

2.7.14 Liquid Waste Disposal Ponds 

The TRA Liquid Waste ponds are east of the ETR reactor building. The 7.5-acre ponds were built for 

the disposal of low-level liquid waste from test reactor operations. When all three test reactors were 

operational, approximately 50 million gallons of wastewater per month were discharged to the 

seepage ponds and the ETR disposal well. Most of the activity pumped to the ponds is 51Cr and 3

H 

(Nebeker and Lakey 1970).

An estimated 3,000 wild ducks per year land on the pond, usually stay less than a week, and have 

some potential to carry activity off the site. An extensive study analyzed the ducks for ingestion of 

134Cs, 137Cs, 75Se, 131I, 239Pu, and 240Pu (RAC 2002).

Internal exposure. The seepage ponds accumulated a significant quantity of mixed fission products. 

The activity was fairly stable as long as water levels remained high. When the water was allowed to 

recede and soil was allowed to dry, the activity could become airborne by the winds.

External exposure occurs to the operators from old accumulated activation products and fission 

products during routine inspection or sampling of the ponds. Several studies of the water and the soil 

in the pond resulted in additional external exposure. 

2.7.15 High-Level Liquid Waste Tanks and Transfer Facility

On the south side of the MTR HP office, there are four 1,500-gallon HLLW catch tanks. Placed 

underground in concrete vaults. Tanks 1 and 2 receive water from the Hot Drain System, which 

includes the MTR floor drains. Tanks 3 and 4 receive liquid waste from the Radiochemistry 

Laboratory and the TRA Hot Cells. The liquid waste is accumulated until the tank is nearly full and 

then it is sampled. If the waste meets low-level waste criteria, it is transferred to the Retention Basin 
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and then to the TRA Liquid Waste Disposal Ponds. If it does not meet the criteria, the waste is 

transferred to the HLLW holding tanks. When the Hot Waste tanks are nearly full, the wastewater is 

loaded on a tank truck and shipped to ICPP for processing. The HLLW tanks consist of two 7,500-

gallon and two 9,000-gallon tanks.

Internal exposure potential exists from airborne radioactivity during entry to the holding tank pits for 

repairs from contaminants of mixed fission products and activation products.

External exposure due to sampling and transferring liquid waste is low because the tanks are in 

concrete-shielded underground vaults. Radiation sources are mixed fission or activation products 

when entry to the pits is necessary. 

2.8 AUXILIARY REACTOR AREA, APRIL 1958 to late 1990s 

The ARA was originally the Army Reactor Experimental Area (AREA), which was changed to the 

Army Reactor Area. It was established as a site to test stationary, portable, or mobile reactors of low, 

medium, or high power units. The ARA reactors were built and maintained by contractors with a 

mixed cadre of military personnel training on the operation of the facilities. ARA is 10 miles east of 

CFA; it began with the ARA-I site one-half mile north of Highway 20. ARA-II, -III, and -IV are at halfmile intervals along an access road, Fillmore Avenue, north from Highway 20. After the Army phased 

out its program around 1965, ARA facilities were used for experiments and tests involving multiple 

radionuclides, particularly at the ARA-I hot cell and laboratory facilities. D&D of the ARA ended in 

2002.

2.8.1 ARA-I 

ARA-I housed a hot cell facility (ARA-25) and laboratory with hoods and metallurgical equipment to 

support reactor and other radiological experimental work. It operated from the early 1960s through 

the late 1990s with periods of inactivity. The Hot Cell was involved with recovery of debris from the 

SL-I excursion accident and associated reactor and fuel experiments. The hot cell and laboratory 

were involved in a wide spectrum of activities from low-level alpha experiments to work with irradiated 

reactor samples exposed to core fluxes, including melted and destroyed fuel assemblies from reactor 

tests. 

2.8.2 ARA-II/SL-1, February 23, 1960, to April 6, 1961

ARA-II, a half mile north of ARA-I, was the site of a low-power, boiling-water reactor designed and 

built by ANL applying its BORAX experience. The reactor, designed to generate only 1000 kilowatts, 

was originally named the Argonne Low Power Reactor. After ANL handed over the finished plant to 

the Army’s operating contractor, Combustion Engineering (CE), the Army named the reactor the 

Stationary Low-Power Reactor Number 1 or SL-1. SL-1 went critical for the first time on August 11, 

1958. It operated for periods between 1 and 6 weeks, and then would be shut down for scheduled 

maintenance and test changes. 

SL-1 was shut down for scheduled annual maintenance on December 23, 1960, and was scheduled 

for a January 4, 1961, startup. During preparation for the run, the reactor went prompt critical at 9:01 

p.m. on January 3, 1961, creating a steam explosion that killed three persons and destroyed the 

reactor. The event released fission products (500,000 Ci in the building, and 1,100 Ci to the 

atmosphere) (Horan and Braun 1993) and created high-level radioactive contamination to 50 rad hr-1

around the ARA-II area. Initial recovery from the accident resulted in short-term exposure exceeding 

500 rad hr-1to personnel in radiation fields. Extensive cleanup efforts followed, including complete 
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dismantlement of the facility. The reactor vessel went to the TAN, some of the contaminated items 

went to the RWMC, and some debris was buried in a specially designated location (two large pits and 

a trench) about 1,600 feet from the SL-1 compound. The walls of the silo, the power conversion and 

fan-floor equipment, the shielding gravel, and the contaminated soil gathered during the cleanup went 

into the pits at SL-1. Three feet of clean earth shielded the material. An exclusion fence with hazard 

warnings around the area remains in place east of the reactor site. Operating power history and 

release information is contained in RAC (2002, p. 117).

2.8.3 ARA-III

ARA-III, another half mile farther north, was the site for the Army Gas-Cooled Reactor Experiment 

(GCRE). GCRE was a water-moderated, nitrogen (gas)-cooled, direct and closed-cycle reactor. It 

generated 2,200 kilowatts of heat, but no electricity. The Army wanted to develop a mobile nuclear 

power plant, and the GCRE was the first phase of that program, proving the principle of this reactor 

concept. The reactor provided engineering and nuclear data for improved components. The GCRE 

was used to train military and civilian personnel in the operation and maintenance of gas-cooled 

reactor systems. 

2.8.4 ARA-IV, March 30, 1961 to May 29, 1964

ARA-IV, another half mile north on Fillmore Avenue, was the site for the Mobile Low-Power Reactor 

(ML-1). The entire ML-1 plant was designed to be transported either by standard cargo transport 

planes or standard Army low-bed trailers in separate packages weighing less than 40 tons each. The 

ML-1 reactor was operated remotely from a control cab about 500 feet away. It could be moved after 

a 36-hour shutdown. The reactor was designed for ease of operation and maintenance by 

technicians at remote installations, for reliable and continuous operation under extreme climatic 

conditions, and for the rigors of shipment and handling under adverse conditions. The Army phased 

out its reactor development program around 1965.

Exposures relate to all of the ARA facilities. 

Internal Exposure potential was typical of that associated with operation of a low-power reactor with 

the exception of the Hot Cell (ARA-I) effluent and the SL-1 excursion, which resulted in fission product 

release (1,100 curies to the atmosphere). Airborne radioactivity consisting of mixed fission products 

and activation products would cause most internal exposure.

External Exposure was significant from the SL-1 excursion and recovery operations. Nine technical 

overexposures resulted, ranging from 15 to 27.3 rem (whole-body dose) (Horan and Braun 1993). 

Hot cell work and D&D efforts in the reactor areas contributed to exposure pathways. Hot cell entries 

after working with irradiated experiments were a source of high exposure. Expected exposure 

potential was from photons greater than 250 keV from the mixed fission products and activation 

products associated with the SL-1, Hot Cell operations, and other work connected with reactor 

operation and maintenance activities. Dosimetry badges were worn by workers in this area. 

However, extremity dosimetry might not have been used by all personnel during recovery from the 

SL-1 accident. 

2.9 WASTE REDUCTION OPERATIONS COMPLEX/POWER BURST FACILITY/SPECIAL 

POWER EXCURSION REACTOR TEST AREAS 

The Waste Reduction Operations Complex (WROC)/Power Burst Facility (PBF)/SPERT area is southcentrally located east of CFA on the INEEL site and 51 miles west of Idaho Falls. The site was 
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originally established to conduct research on small power reactors and reactor safety. Its current 

mission is storage of spent nuclear fuel, treatment and storage of mixed and low-level waste, and 

research to reduce hazardous and mixed waste. As part of the DOE mandate to treat legacy wastes 

and remediate the environment, these facilities now provide safe treatment, storage, and recycling of 

INEEL radioactive, mixed, and low-level wastes. Many of the wastes treated originated at INEEL.

Internal and external exposure will be noted at the end of section 2.9 for all the sub-sections.

2.9.1 SPERT-I, June 11, 1955 to 1964

SPERT-I was an open-tank, light-water-moderated, reflected reactor, originally using 92% enriched 

uranium fuel. The reactor tank, about 4 feet in diameter and 14 feet high, was the first in a series of 

four safety-testing reactors designed to study the behavior of reactors when their power levels 

changed rapidly. Power runaways were deliberately produced by moving the control rods. The 

variables in the thousands of SPERT studies included fuel plate design, core configuration, coolant 

flow, temperature, pressure, reflectors, moderators, and void and temperature coefficients. 

At 12:25 p.m. on November 5, 1962, destructive Test No. 1 was initiated with a plate type core. A 

violent explosion occurred immediately after the final power excursion, during which complete fuel 

plate melting occurred in approximately 8% of the core, with partial melting in approximately 35% of 

the core. It was reported that “those isotopes which were collected were released as gases. No solid 

products were collected.” Test No. 2 began at 8:15 a.m. on November 10, 1963, and Test No. 3 

began at 1:14 p.m. on April 14, 1964 (Miller, Sola, and McCardell 1964). A number of nondestructive 

runs were conducted to gain operational information. All operations were conducted from a control 

building a half mile from the reactor. SPERT-I tests demonstrated the damage-resistant capabilities 

of low-enrichment (4% 235U) uranium-oxide fuel pins similar to those used in water-cooled reactors 

powering large central stations. 

2.9.2 SPERT-II, March 11, 1960, to October 1964

SPERT-II, south and east of SPERT-I, was an extension of the SPERT-I excursion tests. It was a 

closed pressurized-water reactor with coolant flow systems designed for operation with either light or 

heavy water. The pressure vessel was 24-1/2 feet high with a 10-foot inside diameter. Tests with 

heavy water (deuterium, an isotope of hydrogen) were desired because heavy water reactors were of 

growing importance in Canada, Europe, and the United States. In addition, heavy water tests enabled 

verification of physics calculations on the effects of neutron lifetime on power excursions. 

At present, the area is used as a lead storage facility; it has been primarily used for storage since 

1964. PBF-contaminated reactor coolant was stored in a tank at the facility and other components 

were stored in a radioactive material storage area. The radioactive liquid waste and radioactive 

material storage area have been removed.

2.9.3 SPERT-III, December 19, 1958, to June 1968

SPERT-III was the most versatile facility yet developed for studying the inherent safety characteristics 

of nuclear reactors. This reactor (which was planned as the third in the series of SPERT reactors but 

was the second built) provided the widest practical range of control over three variables: temperature, 

pressure, and coolant flow. The reactor was in a pressurized vessel similar to those used in 

commercial power production. Water could flow through the vessel at a rate as high as 20,000 

gallons per minute, handle temperatures as high as 650°F, and pressures as high as 2,500 psi. 
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2.9.4 SPERT-IV, July 24, 1962, to August 1970

SPERT-IV was an open-tank, twin-pool facility that permitted detailed studies of reactor stability as 

affected by varying conditions including forced coolant flow, variable height of water above the core, 

hydrostatic head, and other hydrodynamic effects. The water-moderated and reflected reactor used 

highly enriched, aluminum alloyed, plate-type fuel elements. The SPERT-IV facility was modified by 

the installation of a Capsule Driver Core (CDC), which permitted the insertion of fuel samples in a test 

hole in the center of the reactor core, where they could be subjected to short-period excursions 

without damaging the “driver” fuel in the rest of the core. Work on fuel destructive mechanisms 

continued until the PBF replaced the CDC. 

In commercial plants, the reactor cores contain tons of fuel. Analysts imagined the consequences if 

the coolant somehow failed to carry away the fission heat. Suppose a pipe leaked or broke? The 

SPERT tests had proven that such a situation would easily put a stop to the chain reaction: the loss of 

pressure would allow the water to turn to steam; the lower density of steam would fail to moderate the 

neutrons; and the nuclear reaction would stop. But the radioactive decay of the fission products in the 

fuel elements would continue to produce heat and continue to need cooling. This concern spawned 

the Loss-of-Coolant-Accident Program and the PBF. The SPERT reactors were decommissioned and 

replaced with other operations, as noted below.

2.9.5 Power Burst Facility, September 22, 1972, to 1985

The PBF is a much larger and more sophisticated reactor than the SPERT reactors. It was built on 

the site of the SPERT-I facility. PBF was initially developed to perform tests of nuclear reactor fuels 

during off-normal reactor operations. It was designed to simulate various kinds of imagined accidents 

caused by sudden increases in the reactor operating level. PBF was the only reactor in the world that 

could perform rapid power changes (bursts) within milliseconds. It performed simulated LOCAs and 

severe-fuel-rod-burst tests in a special assembly (loop) in the main reactor core. Fuel damage on 

experiments in the loop would transport fission products throughout the loop piping and through steam 

lines outside the shielded loop cubicle. Monitors detected and timed the precise movement of fission 

products as they escaped from a fuel rod with failed cladding. Data from these tests were used to 

develop and validate fuel behavior computer codes for the NRC. Retrieving data and modification of 

the various test configurations resulted in exposure to high radiation fields and potential for release of 

fission products in the reactor containment. 

The PBF was a high-performance, water-cooled, uranium-oxide-fueled reactor designed to provide 

information on light-water reactors. Airborne effluents were filtered and passed through charcoal beds 

to remove iodine. Liquid wastes were pumped to a disposal well or held in tanks for transport to the 

ICPP. The reactor operated from September 22, 1972, into 1985 until it was placed on standby 

status. In 1998, the PBF was placed in shutdown status and is being prepared for fuel removal.

2.9.6 Lead Storage Facility

The Lead Storage Facility is housed in the old SPERT-II facility. It is used to collect and store clean 

lead for the INEEL emergency lead inventory. The building was used in the past for storage of 

radioactive material. The area is not a radiological concern.

2.9.7 Waste Experimental Reduction Facility, 1982 to 2001

The Waste Experimental Reduction Facility (WERF) is a versatile waste treatment facility that began 

treating low-level radioactive wastes in 1982 at the location of SPERT-III after that facility’s D&D. Its 
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original mission was to reduce the volume of low level radioactive waste through incineration, 

stabilization, compaction, and metal sizing processes to prepare wastes for safe permanent 

disposition prior to burial at the RWMC. In the beginning, metal was sized and melted into ingots in 

two furnaces until it was determined sizing alone was more cost-effective. An incinerator was added 

and used to reduce the volume and increase the stability of a wide variety of low-level wastes prior to 

disposal at the RWMC. In 1984, the incinerator began treating RCRA-defined mixed (radioactive and 

hazardous) waste. Wastes from INEEL and other DOE facilities were treated under provisions in the 

INEEL Site Treatment Plan. INEEL began the closure process for the WERF, which should take 3 

years, in 2001.

2.9.8 Mixed Waste Storage Facility

The Mixed Waste Storage Facility, in the SPERT-IV reactor building, is a RCRA storage facility for 

interim storage of mixed low-level wastes. It has regulatory approval to store polychlorinated 

biphenyls (PCBs), corrosives, and flammables. Treatments are being developed for the types of 

wastes stored in the facility.

Internal exposures were possible based on releases from the various reactor operations at the 

SPERT reactors and PBF. Maintenance activities and other work with radioactive material (especially 

from PBF loop experiment) resulted in airborne mixed fission and activation products, making possible 

internal exposure with 137Cs being the primary radionuclide. 

External exposure resulted from experiment changes related to reactor experiment changes and 

maintenance activities. 137Cs was a primary nuclide for direct radiation exposure from fission products 

in the transport lines and in the loops at the PBF during Severe Fuel Damage tests when radiation 

levels were measured up to 50 rad hr-1 Other radiological work activities resulted in much lower 

exposure rates from the mixed fission products and activation products. 

2.10 ORGANIC MODERATED REACTOR EXPERIMENT, SEPTEMBER 17, 1957, TO APRIL 

1963

The Organic Moderated Reactor Experiment (OMRE), a few miles east of the CFA, was built to test 

the feasibility of the organic-cooled reactor concept. OMRE demonstrated the technical and 

economic feasibility of using a liquid hydrocarbon as both coolant and moderator. The reactor 

operated with a succession of cores. The waxy coolant was considered promising because it 

liquefied at high temperatures but did not corrode metal as water did. In addition, it operated at low 

pressures, significantly reducing the risk of leaks. However, it lacked test loops needed to investigate 

various organic coolants and experimental fuel elements.

A scaled-up reactor, the Experimental Organic Cooled Reactor (EOCR), was built next to OMRE in 

anticipation of further development of the concept. The purpose of EOCR, which had special testing 

loops and other advanced features, was to extend and advance the OMRE studies. During the final 

stages of its construction, EOCR was placed in standby (December 1962) when the AEC decided that 

the organic-cooled concept would not significantly improve performance over what other reactor 

concepts had achieved for nuclear power. EOCR never operated. The building was recycled for 

other (non-nuclear) uses.

Internal Exposure. On November 16, 1960, an experiment was conducted to determine the feasibility 

of open-air burning of contaminated solvents accumulated at the OMRE facility. Approximately 400 

gallons of liquid consisting of diesel oil, xylene, methyl-chloroform, and a small amount of water were 
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placed in an open vessel and ignited. Since the reactor did not operate, no other potential for internal 

dose occurred. 

External exposures were possible from mixed fission products and activation products associated with 

core changes and associated reactor maintenance in radiological areas. 

2.11 TEST GRID III, 1957 TO APRIL 24, 1970

Test Grid III, near Lincoln Boulevard south and east of NRF and east of the TRA and INTEC, was the 

site of several tests with atmospheric releases. It originated in the 1957 timeframe, and was based on 

concerns about what would happen from nuclear aircraft crashes, etc. This highly instrumented grid 

would measure atmospheric conditions and release information of the various tests. (RAC 2002) 

Exposure information is summarized at the end of section 2.11 for the following sub-sections.

2.11.1 Fuel Element Burn Tests

Fuel Element Burn Tests A and B were conducted on test Grid III to support the General Electric ANP 

Program to evaluate the consequences of a nuclear aircraft crash involving a fire. Aged fuel elements 

were heated to assist with understanding the behavior of a fuel element in a large fire and to provide 

initial data on the percentage release of fission products to the environment. Average ground 

radiation levels near the burn site immediately following the burn test were 200 mrem hr-1 (Brodsky 

and Beard 1960). Meteorological conditions had been carefully studied in advance and were closely 

monitored during the tests.

FEBT-A was conducted at 2:19 p.m. on March 20, 1957, using an irradiated fuel element (well-aged) 

containing 5,000 Ci of fission products. A pool of jet fuel was ignited under the fuel element, which 

reached a temperature of about 2,250 ̊F. After the fire, the fuel element was intact with a small 

puncture in the cladding.

FEBT-B used an induction furnace that heated a fuel element containing 10,000 Ci of fission products 

to 5,000  ̊F. Most of the fuel element melted and dispersed within 90 seconds during inversion 

conditions (Brodsky and Beard 1960). The test was conducted at 6:47 p.m. on March 20, 1957. 

FEBT-B was ranked above other release events at both onsite and offsite locations as an episodic 

event in RAC (2002)

2.11.2 Fission Products Field Release Tests, July 25, 1958, to September 26, 1958

Fission Products Field Release Tests (FPFRTs) were conducted on Grid III to represent accidents 

involving nuclear-powered aircraft. Nine tests evaluated release percentages, airborne radioactivity, 

and diffusion and deposition characteristics of fission products released from melted aircraft reactor 

fuel elements (Convair 1959). Five tests were with fuel decayed for 922 to 985 days and four were 

with fuel decayed for 42 to 65 days. Operating temperatures were between 1,000 and 2,300  ̊C. To 

simulate a potential accident, the tests used an induction-type furnace to heat the elements rapidly to 

the melting point in approximately 2 minutes, and maintained this temperature for approximately 10 

minutes after melting began (Convair 1959). Instruments situated about a fan-shaped grid with seven 

concentric arcs and a maximum radius of about 5 miles obtained cloud diffusion, meteorological, 

radiological, radiobiological, and deposition data. RAC (2002) contains information on meteorological 

conditions, furnace temperatures, release fractions, etc.
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2.11.3 Relative Diffusion Tests, November 30, 1967, to October 1, 1969

The four Relative Diffusion Tests (RDTs) involved the intentional release of1 to 6 Ci of both methyl 

and elemental radioiodine. Details on these releases are limited, but some information is in DOE 

(1991b) and RAC (2002). 

2.11.4 Experimental Cloud Exposure Study, May 3, 1968, to April 24, 1970

The Experimental Cloud Exposure Study (EXCES) tests in 1968 and 1969 consisted of 133Xe releases 

ranging from 32 to 600 Ci; tests in 1970 consisted of 24Na releases ranging from 6.6 to 120 Ci. The 

primary objectives for the tests included measuring total exposure at several downwind distances; 

determining dimensions of the plumes; documenting the release rate and height, wind speed, and 

temperature; and measuring the gamma energy spectrum at one or more points during the release. 

Releases were planned to occur during meteorological conditions characterized by winds out of the 

southwest to minimize potential on- and offsite exposure and to ensure that the cloud passed over 

preset instrumentation. Voillequé (1969) discusses an outline of plans for the 133Xe release tests, 

including the general objectives and procedures associated with the tests (RAC 2002). 

Internal exposures were possible from the airborne radioactivity released from the tests and potential 

airborne radioactivity from the materials during handling. Grid III was highly instrumented to detect 

release fractions. The FEBT-B had a high ranking for INEEL releases, and the least amount of 

monitoring, according to RAC (2002). Personnel were involved in the Site’s bioassay program.

External exposure would have resulted from working in the proximity of, handling, or transporting 

irradiated test fuel elements with a potential for exposure from photons greater than 250 keV 

consistent with that of other irradiated fuel loading and unloading tasks. 

2.12 EXPERIMENTAL DAIRY FARM/EXPERIMENTAL FIELD STATION

The EFS was a 27-acre plot about 7 miles northeast of the ICPP near Grid III. This facility, also 

known as the Dairy Farm, was established to further studies on the pathway of 131I from a release to 

the human thyroid. It included pastures, a barn, six cows, and a grid of detection instruments in the 

pasture in regular lines and rows.

2.12.1 Controlled Environmental Radioiodine (Release) Tests, May 27, 1963, to December 

1977

The primary objectives of the Controlled Environmental Radioiodine Test (CERT) releases were to 

establish relationships between the amounts of radioiodine in different environmental media. 

Specifically, these tests studied relationships between air and soil and vegetation, vegetation and 

milk, and milk and human thyroids. They involved releases of elemental and methyl radioiodine 

ranging in amount from 0.05 to 8 Ci. Most of the releases occurred at the Experimental Dairy Farm. 

Others occurred at the ICPP, ARA, NRF, and CFA areas. In 1968, the name was changed to 

Controlled Environmental Release Test to reflect the release and study of additional radionuclides, 

such as cesium, cerium, potassium, and krypton. Hawley (1964) reports that the CERT No. 1 release 

occurred near ground level over a 30-minute period. Additional information regarding this test series 

is provided by Hawley (1964); Bunch (1966 and 1968), and Zimbrick and Voillequé (1969). Early in 

the test program, the AEC granted permission to six DOE-ID volunteers to be a part of a human 

experiment program. During initial tests these volunteers sat in the field during the release. In later 

tests, after the cows had eaten contaminated grass, the volunteers drank small quantities of milk. The 
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series included 29 experiments, although only a few of the early ones involved human consumption of 

milk.

Internal exposures were possible from all the releases and were intentional during the early phases of 

the tests. Exposures are well documented in the dosimetry records.

External exposure was well below the level acceptable for radiological work from the tests. Personnel 

were required to wear dosimetry devices.

2.13 INEEL RESEARCH CENTER, 1984 to present

The INEEL Research Center (IRC), a 35-acre site on North Boulevard in Idaho Falls, was built 

between 1982 and 1984. The facility was dedicated in 1984 to further the INEEL research and 

engineering mission. The IRC has 58 laboratories in IF-603 for geophysics, chemistry, microbiology, 

and other sciences; 18 of these are general-purpose modules for electronics design, optics, lasers or 

materials testing, and nondestructive examination research and development. The Center conducts 

laboratory work with tracer-level radionuclides. One of the early missions was work for the Bureau of 

Mines with ores, some containing natural uranium. The INEEL Engineering Demonstration Facility 

(IF-657) houses several prototypical-scale R&D projects that support programs in military munitions 

assay, advanced sensor systems, environmental restoration, subsurface investigation, and materials 

science. There is a shielded 252Cf source (initially 2.5 mCi) in the high bay of IF-638. The onsite 

radiological control technician has additional low energy Pu, Sr, and Am sources for portable survey 

instrument response checks. 

Internal exposure potentials are minimal because of the radiotracer-level, low-energy alpha and beta 

source material such as natural uranium, 14C, etc.

External exposure. Low-level personnel exposures have been measured from the neutron source (2-

to-20-MeV range) and X-ray (30-to-250-keV range) equipment. Personnel dosimetry is required for all 

work in radiation areas. 

2.14 ARMY REENTRY VEHICLE FACILITY SITE (OR STATION), 1965 TO MAY 1996

The Army built the Army Reentry Vehicle Facility Site (ARVFS) 12 miles northeast of CFA in 1965 for 

classified DOD experiments with an advanced reentry vehicle fuzing system. The facility consisted of 

an open-top cylindrical test pit, an underground bunker, and a system of cables and pulleys between 

the bunker and tank (Thiel 1997; Mobley 1987). 

The ARVFS Facility was used in 1965 to conduct an irradiation study using four spent MTR fuel 

elements to evaluate the accuracy of the Radiological Safety Analysis Computer (RSAC)-generated 

code cloud-gamma exposure information. Movement of the fuel to conduct the experiment and the 

transportation of the fuel to and from the facility was an external exposure source. Dose rates are not 

available (McCaslin 1968).

The bunker was used to store NaK coolant radiologically contaminated from the EBR-I Mark II nuclear 

reactor core meltdown in November 1955. The NaK was stored from 1974 through 1995, when it was 

shipped to ANL-W for reprocessing. Engineering evaluations from 1986 through 1992 provided 

radiological data with maximum radiation exposure rates from the NaK containers to be ~ 40 Rhr-1. 

The radiological fission product inventory in the NaK reported in LaRue and Dolenc (1986) was 23.3 

grams and about 133 Ci. The Final Safety Analysis Report indicates that conservative total dose to 
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process the four NaK containers would be 0.394 rem, assuming the same person was involved with 

each step (Mobley and Keller 1991) 

Internal exposure potential was minimal at ARVFS because airborne radioactivity was not present 

during activities at the bunker.

External exposure occurred during radiological surveys and loading and unloading of the NaK 

containers for storage and or transport. 
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Table 2-1. INEEL prime site contractors. 

Operating years Prime contractor

1950–1966 Phillips Petroleum Company

1966–1972 Idaho Nuclear Corporation (Allied Chemical Corporation, Aerojet General Corporation, 

and Phillips Petroleum Company)

1972–1976 Aerojet Nuclear Corporation

1976–1994 EG&G Idaho

1994–1999 Lockheed Martin Idaho Technologies Company

1999–2004 Bechtel BWXT Idaho, LLC

Table 2-2. Other operating facility technical area contractors. 

Argonne National Laboratory–West

1949–2004 University of Chicago

Idaho Chemical Processing Plant (Idaho Nuclear Technology and Engineering Center)

1950–1953 American Cyanamid Company

1953–1966 Phillips Petroleum Company

1966–1971 Idaho Nuclear Corporation

1971–1979 Allied Chemical Corporation

1979–1984 Exxon Nuclear Idaho Company

1984–1994 Westinghouse Idaho Nuclear Company

Aircraft Nuclear Propulsion Program (ANPP)

1952–1961 General Electric

Specific Manufacturing Capability

1983–1986 Exxon Nuclear Idaho Company

1986–1991 Rockwell INEL

1991–1994 Babcock & Wilcox Idaho, Inc.

Table 2-3. Radionuclides of concern for all reactors. 241Am 60Co 131I 103Ru

140Ba 51Cr 133I 106Ru

141Ce 134Cs 140La 90Sr/90Y

143Ce 137Cs 54Mn 91Sr

144Ce 152Eu 95Nb 92Sr

242Cm 154Eu 239Np 234U

244Cm 59Fe 238Pu 65Zn

58Co 3

H 239Pu 95Zr
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Table 2-4. ICPP 601/602 process cell information. 

Cell Process description Cell function 

A EBR feed preparation Fuel dissolution

C MTR feed preparation Fuel dissolution

D MTR feed preparation Fuel dissolution

E Str-sir feed preparation Fuel dissolution

F First cycle extraction Uranium separation

G MTR feed preparation Fuel dissolution

H MTR first cycle extraction Uranium separation

J Hot salvage

K Solvent recovery

L RaLa process cell Recovery of RaLa 

M

P First cycle extraction

Q Second cycle extraction

R Product transfer cell

S Third cycle extraction

T Solvent pumps Hexone storage

U First cycle aqueous raffinate treatment

V Health physics field office

W First cycle solvent raffinate

X Sample dilution & decontamination

Y Second & third cycle raffinate treatment

Z Product room
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Processing Plant, TD898.2 W32, Idaho Nuclear Corporation (Aerojet), Idaho National 

Engineering and Environmental Laboratory, Idaho Falls, Idaho.

Baca, J. P., 1979, History and Status of the Fuel Examination Facility Complex, Proceedings of 27th 

Conference on Remote Systems Technology, San Francisco, California.

Becar, N. J., J. F. Kunze, G. D. Pincock, and L. D. Vavleck, 1961, Critical Experiment Tank (CET) 

Reactor Hazards Summary, Aircraft Nuclear Propulsion Department, XDC 61-4-709, Idaho 

Falls, Idaho.

Bonney, R. F., M. D. Calley, D. A. Johnson, S. D. Novak, and S. L. Zeigler, 1995, Hazards 

Assessments for Test Area North, INEL-94/0140, Idaho National Engineering and 

Environmental Laboratory, LIMTCO, Idaho Falls, Idaho.

Brittan, R. O., B Cerutti, H. V. Lichtenberger, J. K. Long, R. l. McVean, M. Novick, R. Rice, and F. W. 

Thalgott, 1961 Hazard Evaluation Report on the Fast Reactor Zero Experiment ZPR III, ANL6408, October Idaho Falls, Idaho,.

Bunch, D. F., (ed.), 1966, Controlled Environmental Radioiodine Tests, Progress Report Number Two, 

IDO-12053, August, Idaho Falls, Idaho.

Bunch, D. F., (ed.), 1968, Controlled Environmental Radioiodine Tests, Progress Report Number 

Three, IDO-12063, INEEL, January, Idaho Falls, Idaho.

Brodsky, A. and G. V. Beard (eds.), 1960, A Compendium of Information for Use in Controlling 

Radiation Emergencies, Including Lecture Notes from a Training Session, February 12-14, 

1958, ID-8206, INEEL, September, Idaho Falls, Idaho.

Brunson, G. S., 1959, Design and Hazards Report for the Argonne Fast Source Reactor, AFSR, ANL6024, Idaho Falls, Idaho June.

Casto, W. R., 1980. ICPP Criticality Event of October 17, 1978, Nuclear Safety, volume 21 (5): pp 

648-653, Idaho Falls, Idaho.

Cohen, S. & Associates, Inc., Shonka Research Associates, Woodward-Clyde Consultants,1994, 

Identification, Retrieval, and Evaluation of Documents and Data Pertinent to a Historical Dose 

Reconstruction at the Idaho National Engineering Laboratory, Final Report, Contract No. 200-

92-0538 for the Centers for Disease Control and Prevention, December, Atlanta, Georgia.

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U.S. Air Force Nuclear Aircraft Research Facility, NARF-59-32T, INEEL, September Idaho 

Falls, Idaho.

Cordes, O. L., D. F. Bunch, J. R. Fielding, and J. K. Warkentin, 1967, Radiological Aspects of the 

SNAPTRAN -2 Destructive Test, IDO-17203, INEL, Idaho Falls, Idaho.

Cordes, O. L., J. R. Fielding, G. A. Dineen, and R. P. Bird, 1965, Radiological Aspects of SNAPTRAN 

2/10-3 Destructive Test, INEEL, Idaho Falls, Idaho.
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DOE-ID (U. S. Department of Energy, Idaho Falls Office), 2003, Explanation of Significant 

Differences, Revision 0, for the 674-B pond, DOE-ID 11030, Idaho Falls, Idaho, May.

DOE (U.S. Department of Energy), 1991a, Idaho National Laboratory Historical Dose Evaluation, 

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Evaluation Quality Assurance Files, DOE/ID-12119-QAF-001-128, Idaho Falls, Idaho.

GE (General Electric Corporation), 1961, Idaho Test Station Nuclear Materials and Propulsion 

Operation, DCL-61-11-706, Idaho Falls, Idaho. 

ERDA (Energy Research and Development Administration), 1977, Final Environmental Impact 

Statement, Waste Management Operations, ERDA-1536, INEEL, September, Idaho Falls, 

Idaho.

Freund, G. A., P. Elias, D. R. MacFarlane, J. D. Geier, and J. F. Boland, 1960, Design Summary 

Report on the Transient Test Reactor Facility TREAT, ANL 6034, June, Idaho Falls, Idaho.

Haroldson, R. O., F. D. McGinnis, M. Novick, R. R. Smith, and E. W. Thalgott 1963, Safety Analysis 

Report EBR-1, Mark I, ANL 6411, February, Idaho Falls, Idaho.

Hawley, C. A., Jr. (ed.), 1964, Controlled Environmental Radioiodine Tests at the National Reactor 

Testing Station, IDO-12035, INEEL, June, Idaho Falls, Idaho.

Hoefer J. A, 1957, Summery Hazards Report for Critical Experiments with the HOTCE Reactor 

General Electric APEX-345, Idaho National Engineering and Environmental Laboratory, Idaho 

Falls, Idaho.

Hoff, D. L., E. W. Chew, and S. K. Rope, 1986, 1985, Environmental Monitoring Program Report for 

the Idaho National Engineering Laboratory Site, DOE/ID-12082(85), May, Idaho Falls, Idaho.

Horan, J. R., and J. B. Braun, 1993, Occupational Radiation Exposure History of Idaho Field Office 

Operations at the INEL, EGG-CS-11143, EG&G Idaho, Inc., October, Idaho Falls, Idaho.

INEEL (Idaho National Engineering and Environmental Laboratory), 2003, Design and Waste 

Acceptance Criteria for the INEEL CERCLA Disposal Facility, Idaho Falls, Idaho

INC (Idaho Nuclear Corporation), 1969, Standard Reactor Operating and Maintenance Procedures for 

the Spherical Cavity Reactor Critical Experiment at the Low Power Test Facility CI 1156, Idaho 

National Engineering and Environmental Laboratory, Test Area North (TAN) Area, Idaho Falls, 

Idaho. 

Kitttel J. H., M. Novick, and R. F. Buchanan, 1957, The EBR-I Meltdown - Physical and Metallurgical 

Changes in the Core, ANL-5731, Idaho Falls, Idaho, November.

Koch, L. J., H. O. Monson, D. Okrent, M. Levenson, W. R. Simmons, J. R. Humphreys, J. Haugsnes, 

V. C. Jankus, W. B. Loewenstein, 1957, Hazard Summary Report Experimental Breeder 

Reactor II (EBR-II), ANL 5719, May, Idaho Falls, Idaho.
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Kunze, J. F and P. L. Chase, 1970, Facility Operating Manual, Idaho Nuclear Program for the US 

AEC, Idaho National Engineering and Environmental Laboratory, Test Area North (TAN) Area, 

Idaho Falls, Idaho.

LaRue, D. M., and M. R. Dolenc, 1986, Decontamination and Decommissioning Plan for Processing 

Contaminated NaK at the INEL, EGG-WM-7375, September, Idaho Falls, Idaho

Lofthouse J. H., 1971, Standard Operating and Maintenance Procedure for the Split Table Reactor at 

the Low Power Test Facility, Idaho Nuclear Corporation for the USAEC, CI-1216, Idaho Falls, 

Idaho.

McCaslin, J. W., 1968, Health and Safety Branch Progress Report, Idaho Nuclear Corporation, 

August 1968 (IN-1205l), Idaho Falls, Idaho (TID-4500).

McVean, R. L., G. S. Brunson, B. C. Cerutti, W. B. Loewenstein, F. W. Thalgott, and G. K. Whitham, 

1962, EBR-II Dry Critical Experiments, ANL 6462, Idaho Falls, Idaho, February.

Miller, R. W., A. Sola, and R. K. McCardell, 1964, Report of the SPERT I Destructive Test Program on 

an Aluminum, Plate-Type, Water-Moderated Reactor, IDO-16883, INEEL, June, Idaho Falls, 

Idaho.

Mobley, E. V., 1987, Preliminary Safety Analysis Report for the Decontamination and 

Decommissioning of the ARVFS NaK, EGG-WM-7802, September, Idaho Falls, Idaho.

Mobley, E. V., and D. E. Keller, 1991, Informal Report, Final Safety Analysis Report for the 

Decontamination and Decommissioning of the ARVFS NaK, EGG-WM-9150, Revision 1, 

August, Idaho Falls, Idaho.

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Releases of Radionuclides to Atmosphere, Surface Water, and Ground, NCRP Report 

Number 123, January, Bethesda, Maryland.

Nebeker, R. L. and L. R. Lakey, 1970, A Study of Possible Improvements in Liquid Waste 

Management at the NRTS Test Reactor Area, CI-1173, Idaho Nuclear Corporation National 

Reactor Testing Station, INEEL, April, Idaho Falls, Idaho.

NIOSH (National Institute for Occupational Safety and Health), 2002a, “External Dose Reconstruction 

Implementation Guidelines,” Revision 1, OCAS-IG-001, Office of Compensation Analysis and 

Support, Cincinnati, Ohio.

NIOSH (National Institute for Occupational Safety and Health), 2002b, “Internal Dose Reconstruction 

Implementation Guidelines,” Revision 0, OCAS-IG-002, Cincinnati, Ohio.

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Operations Office), Idaho Falls, Idaho, 1 April.

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Radionuclide Releases from the Idaho National Engineering and Environmental Laboratory, 

Centers for Disease Control and Prevention Department of Health and Human Services, 

October, Idaho Falls, Idaho.
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Richards, W. J. and G. C., McClellan, 1979, “Neutron Radiography at the Hot Fuel Examination 

Facility,” Proceedings of 27th Conference on Remote Systems, November, San Francisco, 

California.

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Facility, EG&G Idaho, EGG-M-11845, Idaho National Engineering and Environmental 

Laboratory, Idaho Falls, Idaho. 

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Environmental Laboratory 1949 - 1999, Department of Energy, Idaho Operation Office, Idaho 

Falls, Idaho.

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(ARVFS) Bunker B17-702, Final Report, INEL/EXT-97-00469, June, Idaho Falls, Idaho. 

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Nuclear Propulsion Program, AEC Research and Development Report. APEX-901, Idaho 

Falls, Idaho. 

Voillequé, P. G., 1969, Outline of Plans for EXCES Xe-133 Releases Tests, INEEL, 13 August, Idaho 

Falls, Idaho.

Walsh W. P, 1961, Test Program and Procedures for STP Program A, General Electric DC-61-7-706, 

Idaho Falls, Idaho.

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Environmental Radioiodine Tests at the National Reactor Testing Station, Progress Report 

Number Four, IDO-12065, INEEL.
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GLOSSARY

absorption

The process by which radiation imparts some or all of its energy to any material through which 

it passes. 

activation

The process of inducing radioactivity by irradiation.

AmBe

A common neutron source created by an alpha particle from 241Am interacting with beryllium to 

produce a large neutron yield with low gamma-ray yield.

Albedo dosimeter

A device to detect and measure slow neutrons generated by higher energy neutrons incident 

on the body and that reflect back into the dosimeter. 

annual dose equivalent

The dose equivalent received in a year. The annual dose equivalent is expressed in units of 

rem (sievert).

anti-Cs

Anticontamination clothing, referring to special clothing worn by personnel for protection from 

radiological contamination.

Atomic Energy Commission

An agency established by the U.S. Government for oversight of nuclear weapons and power 

production; a predecessor to the U.S. Department of Energy.

background radiation

The radiation in an ambient environment which includes cosmic rays, radiation from natural 

sources, and man-made sources.

 

beta (β) dose

A designation for external dose referring to the dose from less-energetic beta, X ray, and/or 

gamma radiation; typically a shallow dose or dose to the lens of the eye. 

beta radiation

Radiation consisting of electrons or positrons emitted at high velocity from the nuclei of certain 

radioactive elements. Most direct fission products emit beta radiation. 

blowdown

Sudden depressurization from a break in a pipe containing pressurized water in a reactor 

system. 

boiling-water reactor

A nuclear reactor concept in which the coolant, water, is permitted to boil as it absorbs the 

heat of the nuclear reaction. 
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breeder reactor

A nuclear reactor concept in which the operation produces a net increase in fissionable reactor 

fuel. 

calcine

The dry solid (grainy or granular) product of a chemical process of removing liquids from a 

solution; also, the process for creating the chemical reaction that removes liquids from a 

solution.

cladding

The outer layer of material encasing a reactor fuel element (e.g., aluminum or zirconium). 

Cladding promotes the transfer of heat from the fuel to the coolant and contains fission 

products and activation products that result from the fissioning of the fuel.

 

collective dose equivalent

The sum of the dose equivalents of all individuals in an exposed population. Collective dose is 

expressed in units of person-rem (person-sievert).

collective effective dose equivalent

The sum of the effective dose equivalents of all individuals in an exposed population. 

Collective effective dose is expressed in units of person-rem (person-sievert)

containment building

A safety feature of reactors typically engineered to be an airtight building, to prevent the 

release of radioactive gases or radiological contamination to the atmosphere or area outside 

the containment.

control rod

A device manipulated within a nuclear reactor constructed of material to absorb neutrons for 

the purpose of slowing down or increasing the nuclear reaction. 

contamination, radioactive

Particulate matter on surfaces or in the air that is radioactive. 

control room

The operating center of a nuclear reactor from which the reactor is operated and monitored.

coolant

In a reactor, a gas or fluid (water, liquid metal, etc.) contacting reactor fuel for the purpose of 

reducing or removing heat generated by the nuclear reaction.

core

That part of the reactor consisting of the fuel and some of the control elements for reactor 

operation.

criticality

A reaction capable of sustaining a nuclear chain reaction.

Curie

A special unit of radioactivity equal to 3.7x1010 disintegrations per second (dps).
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D&D

Decontamination and decommissioning; dismantling or demolishing a facility after removal of 

radioactive and other potentially hazardous materials from the structure and premises. 

decontaminate

A process removing radioactive particles from a person, place, or object.

depleted uranium

Uranium nuclide that has undergone a process to remove 235U resulting primarily in 238U.

dose equivalent (H)

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The dose unit is the rem.

dose

A specific amount of energy from ionizing radiation or a toxic substance absorbed per unit of 

mass.

dose of record

The dose recorded for individual worker files. 

dosimeter

A device used to measure accumulated radiation exposure. 

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation. 

Emergency Core Cooling System

An emergency backup system designed to inject coolant into the core of a reactor to prevent 

the overheating of the fuel and subsequent fuel damage.

enriched uranium

Uranium enhanced from its natural state to contain a higher abundance of the isotope 235U. 

excursion 

Planned or accidental increase in the normal operating power level of nuclear reactions.

exposure

Technically, a measure of X-ray or gamma radiation capability to ionize air (units of Roentgen). 

extremity

The arm from and including the elbow through the fingertips and the leg extending from and 

including the knee through the toes.

film

Generally means a "film packet" that contains one or more pieces of film in a light-tight 

wrapping. The developed film has an image caused by radiation that can be measured using 

an optical densitometer.
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fission

A nuclear transformation characterized by the splitting of a nucleus into at least two other 

nuclei and the release of a relatively large amount of energy.

fission product

Elements or compounds resulting from fission.

flux

The intensity of particles (e.g., the number of neutrons passing through a unit area in unit 

time.)

fuel assembly

An arrangement of nuclear fuel and its cladding material into a particular form and shape for 

use in a nuclear reactor. Fuel may be assembled in plates, rods of various diameters, or other 

shapes.

fuel reprocessing

A chemical process, usually involving several steps, that recovers 235 U and other fissionable 

products from spent fuel.

gamma rays

Short wave length electromagnetic radiation (photons) originating in atomic nuclei and 

accompanying many nuclear reactions (e.g., fission, radioactive decay, and neutron capture) 

in an energy range of 10 keV to 9 MeV.

Gray (Gy)

The special name for the SI unit of absorbed dose (1 Gy = j kg-1).

half-life

The time it takes for one-half of any given number of unstable atoms to decay (disintegrate).

HEU

Highly enriched uranium.

 

hot cell

A specialized shielded laboratory in which radioactive materials may be handled with the aid of 

remotely operated manipulators. The walls and windows of the laboratory are made of 

materials designed to protect workers from radiation. 

hot run

An operational (or test) run of a chemical process and equipment using radioactive materials.

interim storage

A concept in the management of nuclear waste in which the waste is moved to an intermediary 

location between its point of origin and its ultimate storage location.

ionization chamber

A device used to measure exposure or radiation dose. 
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ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

irradiate

To expose a substance to radiation. 

isotope

Nuclides having the same number of protons in their nuclei (same atomic number), but having 

a differing number of neutrons (different mass number).

megawatt

A measure of electrical power equal to 1 million watts. 

meltdown

The melting of nuclear reactor fuel caused by a failure of the coolant to adequately carry away 

heat.

millirem

A unit of radiation equal to one-thousandth of a rem (see rem).

microcurie

A measure of radioactivity equal to one-millionth of a curie. 

mixed waste

Waste that is both chemically hazardous and radioactive. 

moderator

A material used in a nuclear reactor to reduce the natural speed of neutrons ejected from 

fissioning atoms (water, graphite etc.).

natural uranium

Uranium occurring in nature that has not been through an enrichment process.

neutron

A basic particle in a nuclear reaction, electrically neutral, with nearly the same mass as a 

hydrogen atom.

neutron, fast

Neutrons with energy equal or greater than 10 keV.

neutron, thermal

Strictly, neutrons in thermal equilibrium with surroundings. Generally, neutrons with energy 

less than about 0.5 eV.

nuclear energy

Energy released from a nuclear fission or fusion reaction. 

nuclear power plant

An electrical generating facility using nuclear fuel 
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nuclear waste

A general term used for the byproduct nonuseable material resulting from nuclear reactions, 

including high-level, intermediate, low-level, mixed and transuranic waste.

nucleus

That part of an atom consisting of the total positive electrical charge and most of the mass. 

pocket dosimeters

A type of ionization chamber used by personnel to measure radiation dose. Other names are 

pencil dosimeter, pocket pencil, pocket ionization chamber (PIC).

personal dose equivalent Hp(d)

Represents the dose equivalent in soft tissue below a specified point on the body at an 

appropriate depth d. The depths selected for personnel dosimetry are 0.07 mm and 10 mm, 

respectively, for the skin and body. 

photon

A quantum of electromagnetic energy often referred to as X-rays or gamma rays, but also 

including light and radiant heat. 

Pressurized-water reactor

A concept in which water used to cool the reactor core is pressurized to prevent boiling. Heat 

is typically transferred from a primary system to a secondary system.

primary loop

A closed experimental system through which coolant flows as part of the control for a nuclear 

reaction using the main reactor as the primary source for neutron flux. 

proton

An elementary atomic particle with a positive electrical charge equal numerically to the charge 

of the electron and a mass slightly greater than 1 mass unit.

Quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose.

rad

The unit of absorbed dose.

radiation

Energy transferred through air or some other media in the form of particles or waves (see 

ionizing radiation).

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

radioactive waste

Byproducts of nuclear processes that are radioactive and have no useful recyclable purpose 

(see nuclear waste).
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radionuclide

A radioactive species of an atom characterized by the constitution of its nucleus specified by 

the number of protons, neutrons, atomic number, and mass number. 

RaLa

Radioactive lanthanum, one of the fission products of a nuclear reaction; a lanthanum 

recovery process at the INTEC for development weapons.

reactor vessel

A steel container enclosing the fuel elements, control elements, coolant piping, and other 

structures that support the core of a nuclear reactor.

reflector

Part of the structure of some nuclear reactors designed to reflect neutrons back into the core 

of the reactor.

relative biological effectiveness (RBE)

A ratio of the absorbed dose of a reference radiation to the absorbed dose of a test radiation 

producing the same biological effects, other conditions being equal.

rem

A unit of dose equivalent, equal to the product of the rad absorbed and the quality factor.

retention basin

An outdoor basin (of any of several designs) in which liquid solutions are deposited and held 

pending evaporation or the precipitation of solids. 

Roentgen (R ) 

A unit of exposure to gamma (or X-ray) radiation, defined precisely as the quantity of gamma 

(or X) rays that will produce a total charge of 2.58 x 10-4 coulomb in 1 kg of dry air STP. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher (~>100 keV) energy photons.

secondary loop

The system in a reactor receiving transfer heat from a primary system. The secondary system 

is not exposed to the reactor core and is free of radioactivity under normal circumstances. 

semiscale

The informal name of a scale model of a nuclear reactor operated at INEEL. The core 

simulated the heat of a nuclear reaction by electrical means and was used to study the 

behavior of water and steam in accidents involving the loss-of-coolant caused by a break in a 

coolant system.

scram

The sudden shutdown of the nuclear reaction in a reactor, typically caused by rapid insertion 

of control/safety rods, when a dangerous or undesirable condition occurs.

shielding

Any material or obstruction that absorbs (or attenuates) radiation to protect personnel or 

materials from radiation.
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Sievert (Sv)

The SI unit for dose equivalent. (1 Sv = 100 rem.)

skin dose

Absorbed dose at a tissue depth of 7 mg cm-2 ~ 0.07 mm in tissue.

spent nuclear fuel

Reactor fuel containing fission and activation products that can no longer economically sustain 

a chain reaction.

spent fuel storage basin

A pool or pit made of reinforced concrete containing water and used to store spent nuclear 

fuel. The water acts as shielding and as a coolant 

thermoluminescence

Property of a material resulting in light emission caused from excitation from heat.

thermoluminescent dosimeter (TLD)

A device containing solid chips of material that when heated release the stored energy as light. 

The measurement of this light provides a measurement of absorbed dose. 

transuranic (TRU) waste

Contaminated waste materials with nuclides having an atomic number greater than 92, a halflife over 20 years and concentration greater than or equal to 100 nCi gram-1. 

TRIGA 

An acronym for a type of training research reactor build by General Atomics

tritium

A radioactive isotope of hydrogen having one proton and two neutrons. Tritium gas is 

produced in nuclear reactors and used to boost the explosive power of most modern nuclear 

weapons. It is also a constituent of irradiated water associated with reactor operations.

uranium-235

A fissionable uranium isotope enriched for use as a fuel in nuclear reactors

uranium-238

The most common uranium isotope, typically nonfissioning; can be irradiated in a reactor and 

transformed to plutonium for use as fuel.

uranium oxide

A metallic compound of uranium and oxygen useful as nuclear fuel because it has a higher 

melting point than metallic uranium. However, its heat transfer properties are not as efficient 

as those of metallic uranium.

water-moderated reactor

A reactor concept designed so that water slows down the speed of neutrons from fissioning 

atoms. 

waste storage tank

A holding tank for liquid or gaseous wastes which might or might not be radioactive.
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-TKBS-0007-2 Page 70 of 70

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1000 mg/cm2

); used to 

refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin with energies les than 250 keV or 

a radiograph .

zirconium

A metallic element highly resistant to corrosion and often used to make cladding for nuclear 

fuel. It is sometimes alloyed in small amounts in the fuel itself.

zero power

Also called low power; operating a reactor to maintain a chain reaction at an extremely low 

power level producing very little heat. Zero power reactors are used as sensitive laboratory 

tools to pretest experimental loadings of test reactors and for other analytical purposes.